ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller & Associates I MJW Corporation

Page 1 of 47

Document Title:

Idaho National Engineering and 

Environmental Laboratory – Occupational 

Internal Dose

Document Number: ORAUT-TKBS-0007-5 

Revision: 01

Effective Date: 01/13/2006 

Type of Document: TBD 

Supersedes: Revision 00 

Subject Experts: Bryce L. Rich and Douglas Wenzel

Document Owner

Approval: Signature on File Approval Date: 01/12/2006

Norman D. Rohrig, TBD Team Leader

Approval: Signature on File Approval Date: 01/13/2006

Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 01/12/2006

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 01/13/2006

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

FOR DOCUMENTS MARKED AS A TOTAL REWRITE, REVISION, OR PAGE CHANGE, REPLACE THE PRIOR 

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0007-5 Revision No. 01 Effective Date: 01/13/2006 Page 2 of 47

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

08/11/2004 00 New Technical Basis Document for the Idaho National Engineering 

and Environmental Laboratory – Occupational Internal Dose. First 

approved issue. Initiated by Norman D. Rohrig.

10/12/2004 00 PC-1 Approved page change revision. Deletes references to rigorous 

radiation program on page 39. Initiated by Norman D. Rohrig. 

Approval:

Norman D. Rohrig, TBD Team Leader 

Signature on File _______________09/30/2004

Judson L. Kenoyer, Task 3 Manager 

Signature on File _______________10/01/2004

Richard E. Toohey, Project Director 

Signature on File _______________10/01/2004

James W. Neton, Associate Director for Science

Signature on File____________ _______________10/12/2004

01/13/2006 01 Approved issue of Revision 01. Revised to change table number 

format, responds to new information from Worker Outreach Effort 

and revises MDA tables. Incorporates 1 year decay time for Al fuel in 

Tables 5-18 and 5-24. Combined “INTEC & Unknown” and “Other 

Area” categories in Table 5-24 for 1971-80 into “All but ANL-W”. 

Training required: As determined by the Task Manager. Initiated by 

Norman D. Rohrig.
Document No. ORAUT-TKBS-0007-5 Revision No. 01 Effective Date: 01/13/2006 Page 3 of 47

TABLE OF CONTENTS

SECTION TITLE PAGE

Acronyms and AbbreviationsAcronyms and Abbreviations ................................................................... 5

5.1 Introduction and Historical Overview ..........................................................................................8

5.1.1 Test Reactors................................................................................................................10

5.1.2 High-Enriched Spent Fuel Chemical Processing ..........................................................10

5.1.3 Other Nuclear Facilities and Processes ........................................................................10

5.1.4 Radioactive Nuclides of Concern and Solubility............................................................11

5.2 Radiological Protection Program Management and Support....................................................13

5.2.1 Internal Dosimetry Issues Related to Contractor Changes ...........................................13

5.2.2 Bioassay Programs .......................................................................................................14

5.2.3 Internal Dose Records ..................................................................................................15

5.2.4 Internal Dose Regulations, Investigation Levels, and Data Codes ...............................16

5.3 Internal Dose Control................................................................................................................18

5.3.1 Air Monitoring................................................................................................................19

5.3.2 Early Technical and Analytical Capabilities at the National Reactor 

 Testing Station..............................................................................................................20

5.4 Minimum Detectable Activities ..................................................................................................23

5.5 Development and Impact of Whole-Body Counting ..................................................................23

5.6 Specific Facilities ......................................................................................................................27

5.6.1 Test Area North.............................................................................................................27

5.6.1.1 Specific Manufacturing Capability Project.......................................................28

5.6.2 Idaho Chemical Processing Plant .................................................................................29

5.6.2.1 High-Enriched Spent Fuel Storage ................................................................. 29

5.6.2.2 High-Level Wastes ..........................................................................................31

5.6.2.3 High-Level Waste Calcination .........................................................................31

5.6.2.4 Process Analytical Facilities............................................................................31

5.6.2.5 Spent Fuel Processing and ICPP Most-Limiting Radionuclides......................31

5.6.3 Argonne National Laboratories-West ............................................................................33

5.6.4 Radioactive Waste Management Complex ...................................................................33

5.6.5 Waste Reduction Operations Complex .........................................................................34

5.6.6 Test Reactor Area.........................................................................................................35

5.6.6.1 Test Reactor Area Reactors............................................................................35

5.6.6.2 Test Reactor Area Laboratories ......................................................................36

5.7 Default for Missed Dose............................................................................................................37

5.8 Unmonitored Workers ...............................................................................................................37

References ..........................................................................................................................................39

Glossary...............................................................................................................................................44
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LIST OF TABLES

TABLE TITLE PAGE

5-1 Recorded internal doses for 1992 to 2000................................................................................11

5-2 Primary radionuclides of concern..............................................................................................12

5-3 Routine bioassay history summary ...........................................................................................14

5-4 Internal dose assessment information after 1989.....................................................................16

5-5 Internal dose assessment information before 1989..................................................................16

5-6 Analytical information that could be in claimant dose files........................................................17

5-7 Area codes that could be in claimant dose files........................................................................17

5-8 Derived investigation levels in 1977 for acute exposures .........................................................18

5-9 Surface contamination control and MDAs.................................................................................19

5-10 Urinalysis results in 1959, 1960, and 1961 ...............................................................................22

5-11 Detection limits applicable to environmental sample analyses (1953 to 1965).........................23

5-12 MDAs for urine samples by period............................................................................................24

5-13 MDAs for fecal samples by period ............................................................................................25

5-14 In Vivo MDAs by period ............................................................................................................25

5-15 Summary statistics from the 1963 WBC program.....................................................................27

5-16 Mass and activity ratios of SMC DU uranium isotopes .............................................................28

5-17 Notable airborne incidents at ICPP...........................................................................................30

5-18 Radiologically significant radionuclides for ICPP processed fuels............................................31

5-19 Radionuclides of concern and MDAs for ANL-W locations.......................................................33

5-20 Radioactive waste inventory in the TSA ...................................................................................34

5-21 Radioactive low-level waste inventory in the active pits in the SDA .........................................34

5-22 Radiologically significant radionuclides for ATR fuel gaseous releases ...................................36

5-23 Major airborne incidents at TRA ...............................................................................................36

5-24 Default table for missed dose ...................................................................................................38

LIST OF FIGURES

FIGURE TITLE PAGE

5-1 History of Urine Bioassay at the INEEL. ...................................................................................21
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ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

AEDE annual effective dose equivalent

AMAD activity median aerodynamic diameter

ANL-W Argonne National Laboratory-West

ANP aircraft nuclear propulsion

ANSI American National Standards Institute

ATR Advanced Test Reactor

CAM continuous air monitor

CDE committed dose equivalent

CEDE committed effective dose equivalent

CFA Central Facilities Area

C. F. R. Code of Federal Regulations

Ci curie 

COO Chicago Operations Office

cpm counts per minute

CPP Chemical Processing Plant

d day

DAC derived air concentration

DOE U.S. Department of Energy

DOELAP DOE Laboratory Accreditation Program

dpm disintegrations per minute

DU depleted uranium

EBR-I Experimental Breeder Reactor No. 1

EBR-II Experimental Breeder Reactor No. 2

ERDA Energy Research and Development Administration

ETR Engineering Test Reactor

F fast absorption type

FCF Fuel Cutting Facility

gal gallon

H&S Health and Safety

HFEF Hot Fuel Examination Facility

hr hour

HSD Health and Safety Division

HSL Health Services Laboratory

ICPP Idaho Chemical Processing Plant

ICRP International Commission on Radiological Protection

IDO Idaho Operations Office

IET initial engine test

INEEL Idaho National Engineering and Environmental Laboratory

INTEC Idaho Nuclear Technology and Engineering Center

L liter
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m meter

M moderate absorption type

MAP mixed activation product

mCi millicurie

MDA minimum detectable activity or amount

MDL minimum detectable level

MeV megavolt-electron, 1 million electron volts

MFP mixed fission product

mg milligram

min minute

ml milliliter

MPBB maximum permissible body burden

MPC maximum permissible concentration

MPCa MPC for airborne activity

mrem millirem

mrep millirep

MTR Materials Testing Reactor 

NaI(Tl) sodium iodide doped with thallium

nCi nanocurie

NIOSH National Institute for Occupational Safety and Health

NRF Naval Reactors Facility

NRTS National Reactor Testing Station

PBF Power Burst Facility

pCi picocurie

PCS primary coolant system

RaLa radioactive lanthanum

RAM radiation area monitor or remote area monitor

RCIMS Radiation Control Information Management System

RDR Radiation Dosimetry and Records

RESL Radiological Environmental Sciences Laboratory

RWMC Radioactive Waste Management Complex

S slow absorption type

SDA Subsurface Disposal Area

SL-1 Stationary Low-Power Reactor

SMC specific manufacturing capability

TAN Test Area North

TBD technical basis document

TLV threshold limit value

TMI Three Mile Island

TRA Test Reactor Area

TRU transuranic 

TSA Transuranic Storage Area

U.S.C. United States Code

WBC whole-body counting

WERF Waste Experimental Reduction Facility
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yr year

ZPPR Zero Power Plutonium Reactor (later Zero Power Physics Reactor)

α alpha particle

β beta particle 

γ gamma

σ standard deviation

μCi microcurie 

μg microgram

μm micrometer

§ section
Document No. ORAUT-TKBS-0007-5 Revision No. 01 Effective Date: 01/13/2006 Page 8 of 47

5.1 INTRODUCTION AND HISTORICAL OVERVIEW

Technical basis documents and site profile documents are general working documents that provide 

guidance concerning the preparation of dose reconstructions at particular sites or categories of sites. 

They will be revised in the event additional relevant information is obtained about the affected site(s). 

These documents may be used to assist the National Institute for Occupational Safety and Health 

(NIOSH) in the completion of the individual work required for each dose reconstruction.

In this document the word “facility” is used as a general term for an area, building, or group of 

buildings that served a specific purpose at a site. It does not necessarily connote an “atomic weapons 

employer facility” or a “Department of Energy [DOE] facility” as defined in the Energy Employees

Occupational Illness Compensation Program Act [EEOICPA; 42 U.S.C. § 7384l(5) and (12)]. 

EEOICPA defines a DOE facility as “any building, structure, or premise, including the grounds upon 

which such building, structure, or premise is located ... in which operations are, or have been, 

conducted by, or on behalf of, the Department of Energy (except for buildings, structures, premises, 

grounds, or operations ... pertaining to the Naval Nuclear Propulsion Program)” [42 U.S.C. § 

7384l(12)]. Accordingly, except for the exclusion for the Naval Nuclear Propulsion Program noted 

above, any facility that performs or performed DOE operations of any nature whatsoever is a DOE 

facility encompassed by EEOICPA.

For employees of DOE or its contractors with cancer, the DOE facility definition only determines 

eligibility for a dose reconstruction, which is a prerequisite to a compensation decision (except for 

members of the Special Exposure Cohort). The compensation decision for cancer claimants is based 

on a section of the statute entitled “Exposure in the Performance of Duty.” That provision [42 U.S.C. § 

7384n(b)] says that an individual with cancer “shall be determined to have sustained that cancer in the 

performance of duty for purposes of the compensation program if, and only if, the cancer ... was at 

least as likely as not related to employment at the facility [where the employee worked], as 

determined in accordance with the [probability of causation] guidelines established under subsection 

(c) ....” [42 U.S.C. § 7384n(b)]. Neither the statute nor the probability of causation guidelines (nor the 

dose reconstruction regulation) define “performance of duty” for DOE employees with a covered 

cancer or restrict the “duty” to nuclear weapons work.

As noted above, the statute includes a definition of a DOE facility that excludes “buildings, structures, 

premises, grounds, or operations covered by Executive Order No. 12344, dated February 1, 1982 (42

U.S.C. 7158 note), pertaining to the Naval Nuclear Propulsion Program” [42 U.S.C. § 7384l(12)]. 

While this definition contains an exclusion with respect to the Naval Nuclear Propulsion Program, the 

section of EEOICPA that deals with the compensation decision for covered employees with cancer 

[i.e., 42 U.S.C. § 7384n(b), entitled “Exposure in the Performance of Duty”] does not contain such an 

exclusion. Therefore, the statute requires NIOSH to include all radiation exposures in its dose 

reconstructions for employees at DOE facilities, including radiation exposures related to the Naval 

Nuclear Propulsion Program. As a result, all internal and external dosimetry results are considered 

valid for use in dose reconstruction. No efforts are made to determine the eligibility of any fraction of 

total measured exposure for inclusion in dose reconstruction.

This site profile provides specific information concerning documentation of historical practices at the 

Idaho National Engineering and Environmental Laboratory (INEEL).

In 1949, the U.S. Atomic Energy Commission (AEC) established the National Reactor Testing Station 

(NRTS) and started construction of facilities on a 572,000-acre site approximately 50 miles west of 

Idaho Falls in southeastern Idaho. NRTS was later the Idaho National Engineering Laboratory and 

then the INEEL. In 2005, DOE shortened the name to the Idaho National Laboratory.
Document No. ORAUT-TKBS-0007-5 Revision No. 01 Effective Date: 01/13/2006 Page 9 of 47

Each of the original AEC laboratories was unique in both mission and location. Because the early 

days of the AEC programs represented the beginnings of the nuclear age, significant technical 

developments were a necessity, not the least of which were developments in radiation safety. Some 

of the unique characteristics of radiation safety (and internal dosimetry specifically) at the NRTS that

had a marked influence on the internal dosimetry programs at each of the facilities were:

The original mission of the NRTS was (as the name implied) high-enriched (over 50% and mostly over 

90%) uranium reactor concept development, materials testing through high-flux test reactor operation, 

and chemical processing of those high-enriched (valuable) uranium fuels. The production of 

weapons-grade nuclear materials was not a mission.

The NRTS beginning was 8 to 10 yr later than Oak Ridge National Laboratory and Hanford. During 

those developmental years significant technical progress in professional skills, instrumentation, 

analyses, procedures, and techniques were accomplished. Radiation safety programs and 

techniques from Oak Ridge (ACC 1952) were adopted at startup of the NRTS facilities. 

Two AEC field offices (Chicago and Idaho) were responsible for and had oversight of the INEEL 

programs included in this report. In addition, the Nuclear Navy under direction of the Pittsburgh Field 

Office administered programs and used facilities on the NRTS for training and program development. 

There were thus three Federal organizations utilizing the NRTS and its infrastructure. This TBD does 

not apply to naval facilities or personnel, even if those personnel received exposure from AEC 

operations. 

In addition to the Federal agencies involved at the site, numerous contractors operated the many 

facilities for the agencies and shared some support personnel to varying degrees.

To provide consistency of radiation safety programs at the NRTS among a large variety of facilities 

and constantly changing contractors, the AEC established a Health and Safety (H&S) Laboratory at 

NRTS to provide technical support in the areas of 1) environmental surveillance, 2) external dosimetry 

(personnel dosimeters of all types), 3) portable radiation detection instrumentation inventories, 

calibration, and maintenance, 4) internal in vitro and in vivo bioassay analytical laboratories, 5) 

technical support in quality assurance of external and internal radiation dose evaluation, 6) 

maintenance and documentation of personnel dosimetry records, and 7) research and development in 

these areas of responsibility. The name of this organization changed to Health Services Laboratory 

(HSL), then to Health and Safety Division (HSD), then to Idaho Center for Radiological and 

Environmental Sciences, and most recently to the Radiological and Environmental Sciences 

Laboratory (RESL). 

Although the design and administration of the radiation safety programs in the workplace were the 

responsibility of each facility contractor, AEC conducted oversight. Technical data, information 

(particularly in the instances of detectable worker intake), and analytical internal dose calculations and 

evaluations were exchanged between the AEC HSL and each contractor.

As a consequence, and in spite of the constant changes at the INEEL, basic assumptions about 

minimum detectable activities (MDAs) or minimum detectable levels (MDLs), missed dose potential, 

and so forth are relatively consistent across the years. There were differences in the available 

nuclear materials from facility to facility, but as early as 1955 or 1956 gamma spectral analysis 

capabilities at the NRTS allowed the significant bioassay results (those which would result in 

reportable internal dose) to be defined in terms of the specific radionuclides. The practice in the case 

of a higher urine sample result was to attempt radionuclide identification through gamma spectral 

analysis and chemical separation. This document describes default assumptions for use in cases 
Document No. ORAUT-TKBS-0007-5 Revision No. 01 Effective Date: 01/13/2006 Page 10 of 47

when the bioassay records for a worker do not include the radionuclide analyses and only record 

gross beta or gross alpha results.

When an incident occurred, it was the policy to investigate thoroughly and identity all individuals 

involved in the incident. When there is no evidence in the incident file or the individual’s dosimetry file

that an individual was involved, and no other evidence supporting that an individual was involved in 

the incident, dose reconstructors should assume that the individual was not involved. 

The majority of this document provides background information to aid the internal dose reconstructor 

through increased general understanding, data interpretation, defaults, and so forth. Sections 5.1.1 to 

5.1.3 provide facility descriptions, and Section 5.1.4 details the radionuclides of concern. Section 5.2 

describes the INEEL radiological protection program as it evolved over the years, Section 5.3 

discusses internal dose control, and Sections 5.4 and 5.5 describe MDAs and whole-body counting

(WBC), respectively. Section 5.6 contains specific information necessary for dose reconstruction

when the facility or facilities where the worker worked are known. Sections 5.7 and 5.8 describe the 

treatment of missed dose and unmonitored workers.

5.1.1

NRTS, which was the primary nuclear reactor development laboratory in the United States, tested or 

evaluated more than 100 reactor concepts (DOE 1997). Fifty-two test reactors were designed, 

constructed, and operated (including operation-to-destruction tests) on the NRTS. INEEL has 

experienced a number of episodic reactor events, both planned and accidental [for example, the 

military Stationary Low-Power Reactor (SL-1) accident on January 3, 1961 (Stacy 2000); a series of 

deliberate safety experiments conducted by Argonne National Laboratory-West (ANL-W) in which 

reactors were allowed to go prompt-critical with resultant reactor destruction (Stacy 2000); and the 

Aircraft Nuclear Propulsion (ANP) Program operated initial engine tests (IET) with large environmental 

releases in the 1950s (DOE 1991)]. External and internal doses to workers, both expected and 

accidental, were associated with these events (RAC 2002). 

Test Reactors

The largest internal exposures at INEEL resulted from accidental intakes associated with episodic 

events or planned major releases, for which the time and characterization of the materials of the 

intakes were well known. These exposures were documented in each exposed employee’s file.

5.1.2

In addition to being the site for experimental test reactors, INEEL is the home of the Idaho Nuclear 

Technology and Engineering Center (INTEC), formerly known as the Idaho Chemical Processing 

Plant (ICPP). The ICPP reprocessed high-enriched reactor fuel (235U enrichments of 50% to 93%) for 

39 yr from 1952 to 1991. Aged mixed fission products (MFPs) were the predominant internal hazard, 

although enriched uranium isotopes and plutonium isotopes (238Pu enhanced) were limiting in specific 

process locations. ICPP experienced not only operational containment barrier failures but also had 

accidental criticality events in 1959, 1961, and 1978. Because the criticality accidents occurred in 

process vessels in heavily shielded cells, these events resulted in relatively minor worker intakes 

(Stacy 2000; AEC 1960; AEC 1962a). These exposures are documented in the personnel dosimetry 

files. 

High-Enriched Spent Fuel Chemical Processing

5.1.3

Other nuclear facilities at INEEL that resulted in internal dose potential or experience include: 

Other Nuclear Facilities and Processes
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The Radioactive Waste Management Complex (RWMC) handled radioactive wastes generated by 

nuclear facilities on INEEL and was the primary disposal location for the Rocky Flats Plant. Although 

most waste came to the RWMC in packages, accidents occurred during handling and processing that 

resulted in releases. This in turn caused intakes of aged MFPs, uranium isotopes, transuranic (TRU) 

radionuclides, and aged mixed activation products (MAPs) (Reilly 2001).

The Specific Manufacturing Capability (SMC) Project is a depleted uranium (DU) specialty-parts 

production plant built in 1985 in the ANP Program hanger on the Test Area North (TAN) site. The 

SMC Project processes metric tons of DU metal for the production of military shielding units (Stacy 

2000). The processes of cutting, machining, and handling uranium metal produce environments in 

which both chronic and accidental intakes of DU have occurred.

At the Naval Reactors Facility (NRF), the U.S. Navy used the INEEL for operating reactors and as a 

naval reactor training center. Because this is not a U.S. Department of Energy (DOE) program and 

not under the oversight of DOE, NRF is not part of the dose reconstruction and compensation 

program. However, through the years NRF has participated in limited coordination of radiological 

protection programs and site support services. It is possible that some workers’ internal dose could 

have resulted from their support work at the NRF.

5.1.4

INEEL facilities and activities have related primarily to experimental reactor design and development, 

spent fuel processing, DU parts production, and low- and high-level radioactive waste treatment and 

disposal. The INEEL Site Description TBD (ORAUT 2004) describes these activities in more detail. 

Table 5-1, which lists the internal doses received at the INEEL from 1992 to 2000, demonstrates that 

the radionuclides of more recent concern to internal dose are MFPs (primarily aged), uranium and its 

decay products, and TRU wastes and their decay products. The table lists internal doses as 

committed effective dose equivalents (CEDEs). This information is from Reilly (2001) and can be 

useful to the reconstruction effort primarily as an indicator of intake radionuclide identity and quantities

Radioactive Nuclides of Concern and Solubility

Table 5-1. Recorded internal doses for 1992 to 2000.

Year of dose 

assignment Facility CEDE (mrem) Radionuclides of concern

2000 SMC 10, 31, 20, 10, 13, and 20 U-238

2000 INTEC (ICPP) 11 Pu-238, Pu-239/240

1999 SMC 15, 48, 13, and 12 U-238

1999 INTEC (ICPP) 16 Pu-239/240

1998 SMC 16 U-238

1997 TRA 10a and 10 a Eu-152 and Eu-154

1997 INTEC (ICPP) 24 Pu-238 and Pu-239/240

1997 TAN 13 a Am-241, Cs-137, Sr-90, U-233, U-238, 

Pu-238, Pu-239/240

1997 SMC 16 & 20 U-238

1996 RWMC 43 Pu-239

1996 INTEC (ICPP) 15, 87, 136, 652a, 655a, 677a, and 678a Am-241, Cs-137, Sr-90, Pu-238, Pu-239/240

1996 SMC 10, 10, 12, 16, 17, 18, 20, 20, & 23 U-238

1995 INTEC (ICPP) 10, 13, 14, 15, 23, 28, 29, 42, 45, & 53 Am-241, Cs-137, Sr-90, Pu-238, Pu-239/240

1995 SMC 10, 12, 14, 15, 19, & 26 U-238

1994 INTEC (ICPP) 14, 20, 25, & 29 Am-241, Cs-137, Pu-238, Pu-239/240

1994 SMC 10, 10, 12, & 15 U-238

1993 INTEC (ICPP) 14, 35, 36, 39, 50, & 53 Am-241, Pu-238, Pu-239/240

1993 SMC 11 U-238

1992 RWMC 20 & 205 Am-241, Pu-238, Pu-239/240

1992 SMC 11, 12, 12, 14, 15, 15, 16, 20, 32, & 52 U-238

a. Internal doses determined after a “significant” event when intake occurred. However, even the other intakes were defined as a result of 

contamination incidents or high airborne measurements. 
Document No. ORAUT-TKBS-0007-5 Revision No. 01 Effective Date: 01/13/2006 Page 12 of 47

for this recent 10-yr period. Before 1990, workers received higher doses from the radionuclides in 

Table 5-1, as well as from MAPs (primarily longer-lived, i.e., 110mAg and others). Most internal doses 

have been identified following an incident rather than as a result of routine bioassay measurements.

Table 5-2 lists radionuclides of concern at INEEL from these programs and as documented in INEEL 

M & O Contractor Technical Basis for Internal Dosimetry, General Technical Basis and Facility 

Specific Documents (Reilly 2001) and from HSL/RESL reports. These radionuclides are those for 

which internal doses were determined in the past and/or for which detection methods were developed. 

The INEEL program follows the DOE guidance and uses the International Commission on 

Radiological Protection (ICRP) Publication 30 nomenclature for solubility (ICRP 1979) as noted in the 

table. In addition, Table 5-2 provides the ICRP Publication 68 (ICRP 1995) recommended absorption 

type.

Through the years at INEEL, plotting urine and fecal elimination curves has shown that an effective 

solubility of moderate (M) to slow (S) per the ICRP Publication 66 system (ICRP 1994) is to be 

expected for the radionuclides listed in Table 5-2, with the exceptions of strontium, cesium, iodine, 

mercury, and tritium. Elimination curves for 3

H, 90Sr, 131I, 137Cs, and 203Hg show they are typically of a 

fast absorption type (F). The chemical explanation is that radioactive materials in oxygen 

Table 5-2. Primary radionuclides of concern. 

Element Radionuclides

ICRP 30a

retention 

class

ICRP 68 

absorption 

type

Preferred 

analysis technique

Hydrogen H-3 (assume HTO) D F Urine

Chromium Cr-51 F,M,S WBC

Manganese Mn-54 F,M WBC

Iron Fe-59 F,M WBC

Cobalt Co-58, Co-60 W,Y M,S WBC

Zinc Zn-65 S WBC

Strontium Sr-89, Sr-90/Y-90b D F Urine

Zirconium/niobium Zr/Nb-95 W,Y F,M,S WBC

Molybdenum Mo-99 F,S Urine

Technetium Tc-99 F,M Urine

Ruthenium Ru-103, Ru-106 F,M,S Urine

Silver Ag-110m F,M,S WBC

Antimony Sb-122, Sb-125 F,M WBC

Tellurium Te-132 F,M Urine, WBC

Iodine I-131, I-133 (assume elemental I2) F Urine, thyroid counter

Cesium Cs-134, Cs-137 D F WBC

Barium/lanthanum Ba/La-140 F/F,M WBC

Cerium Ce-141, Ce-144 W,Y M,S WBC

Europium Eu-152, Eu-154, Eu-155 W,Y M WBC

Gadolinium Gd-153 F,M WBC

Tantalum Ta-182 M,S WBC

Mercury Hg-203 (assume inorganic) F Urine, WBC

Protactinium Pa-233 M, S WBC

Uranium U-233, U-234, U-235, U-236, U-238 D,W,Y F,M, S Urine, fecal

Neptunium Np-237 M Urine, fecal

Plutonium Pu-238, Pu-239/240 W,Y M, S Urine, fecal

Americium Am-241 W M Urine, fecal, lung counter

a. Assigned in the INEEL internal dosimetry technical basis document (Reilly 2001); ICRP 30 = ICRP 1979; ICRP 68 = 

ICRP 1995.

b. Yttrium has an absorption type of M or S; however, in the case of short-lived Y-90, it tracks its parent Sr-90, which is 

normally a type F at the INEEL.
Document No. ORAUT-TKBS-0007-5 Revision No. 01 Effective Date: 01/13/2006 Page 13 of 47

atmospheres oxidize rapidly, which forms less soluble compounds. The default assumption of M or S 

would be appropriate when based on the more claimant-favorable result to the organ in question.

5.2 RADIOLOGICAL PROTECTION PROGRAM MANAGEMENT AND SUPPORT

5.2.1

The changes in contractors at INEEL during its 54-yr history [listed in Tables 2-1 and 2-2 of the Site 

Description TBD of this Site Profile (ORAUT 2004)] resulted in relatively frequent management 

changes at most of the facilities. (The contract with the University of Chicago to operate the ANL-W 

facilities has not changed in this 54-yr period.) 

Internal Dosimetry Issues Related to Contractor Changes

The primary AEC/DOE oversight for NRTS/INEEL, which included most projects on the site and all 

support functions, was assigned to the AEC Idaho Operations Office (IDO). The AEC created the 

H&S Laboratory to provide a variety of health and safety support functions to the entire site, which 

included external and internal dosimetry, health physics instrumentation, fire protection, medical 

services, and environmental surveillance. The Chicago Operations Office (COO) provided oversight 

for ANL-W (contracted by the University of Chicago) programs and facilities, while the U.S. Navy 

provided oversight for the NRF program. ANL-W, which is a DOE program, is included in this TBD. 

The ANL-W program uses site support services, including internal dosimetry support, but the bioassay 

results are reported through COO. 

The INEEL personnel dosimetry records have been and are documented and permanently maintained 

by the various organizations over time. Records about individual facility or contractor field monitoring 

programs (air-monitoring data, personnel contamination records, etc.) were maintained by individual 

contractors and/or site areas and are not maintained in a single recordkeeping system. The field 

monitoring data were not available for use in this report. 

In spite of the frequent changes in operational responsibility through the years and the movement of 

workers among facilities, there has been a basic level of consistency in the internal dosimetry 

programs at INEEL, particularly the bioassay analytical techniques and calculation processes. 

Although the field programs (which monitored the workplace and identified work groups to be included 

in the routine bioassay programs and workers who needed special bioassays) were implemented by 

the individual contractors, there was routine interaction with the H&S Laboratory professionals in

interpretation of dosimetry results as well as in determination of necessary corrective practices or 

procedures. 

Employees were typically assigned to individual facilities and were monitored for specific radiological 

hazards associated with the work. During periods when a single prime contractor was responsible for 

programs at most facilities or for site-wide support personnel, workers in certain crafts (e.g., 

maintenance, specialty operators, and some health physics technicians) worked at several facilities 

and were exposed to a variety of radioactive materials in a variety of work situations. 

The internal dose reconstruction for personnel who worked at a number of INEEL facilities should rely 

on specific bioassay data (radionuclides, quantities, etc.) when available. The procedures and 

technical capabilities for collecting and analyzing bioassay samples at the different facilities were 

basically equivalent. In addition, both the individual facilities and the H&S laboratories had 

radionuclide identification capabilities from the early 1960s. Positive bioassay results (analyses in 

which the results exceeded 2σ counting statistics) were normally followed by a confirmatory analysis

to identify specific radionuclides. In the cases where only gross beta or gross alpha bioassay results 

were available, the results were normally within 2σ. If it is necessary to evaluate intakes from the 

gross beta or gross alpha results, the radionuclide defaults should be 90Sr/90Y and 238Pu, respectively. 
Document No. ORAUT-TKBS-0007-5 Revision No. 01 Effective Date: 01/13/2006 Page 14 of 47

5.2.2

Routine bioassay of radiation workers has been conducted since the beginning of site operations. 

However, formal documentation of the bioassay programs was not found for periods before 1981. 

Some of the data sheets on individuals indicate that bioassay sampling occurred routinely every 

6 months in 1953. Table 5-3 lists the reconstructed history of routine bioassay frequency.

Bioassay Programs

Table 5-3. Routine bioassay history summary.a

Year

Typical 

frequency Type Groups analyzed/sampled Investigating level Comments Reference

1953-

1960 

Annually In vitro urine Radiation workers Frequency is inferred 

from individual data 

sheets. 

Individual 

data 

sheets 

Table 5-10 

AEC 1959 

AEC 1961

1961 Annually In vitro urine 

In vivo 

Radiation workers Frequency is inferred 

from individual data 

sheets.

Table 5-10 

AEC 

1962a

1962-

1972 

Annually In vitro urine

In vivo 

Radiation workers Frequency is inferred 

from individual data 

sheets. 

AEC 

1962a 

AEC 

1963a

1973-

1981 

Annually In vitro urine Radiation workers

Annual DE >10% quarterly 

standard in ERDA Manual 

Chapter 0524 (ERDA 1975) 

Reporting Frequency is inferred 

from individual data 

sheets. 

AEC 1968 

AEC 1975 

ERDA 

1975 

When internal 

intake suspected

Fecal

Annually In vivo

1982-

1987 

Annual In vitro urine CPP-603 workers

Fuel reprocessing operators 50-yr CDE >10% quarterly 

standard in ERDA Manual 

Chapter 0524 (ERDA 1975) 

Reporting Staggered to monitor 

group throughout the 

year 

Exxon 

1981 

When internal 

intake suspected 

Fecal Waste reprocessing operators 

Shift lab workers 

Health physics technicians 

Annual In vivo Selected radio-chemistry 

workers

Maintenance workers

Denitrator operators

Termination In vivo All radiation workers

1988-

1989 

1 to 6 months In vivo All radiation workers

Lung, 50-yr CDE 

>0.5–1.0 rem 

Investigating

Bone Surface, 50-yr CDE 

>1.0–2.0 rem 

Other Organs, 50-yr CDE 

>0.5-1.0 rem 

Staggered so that 

worker receives some 

sort of analysis or 

sampling every 3 

months. 

Tschaeche 

Annually In vitro fecal 1988 

18 to 24 months In vitro urine

Termination In vivo When internal exposure 

In vitro suspected. 

New hire In vitro Depending on review of 

In vivo radiation dose history. 

1990-

1994 

Annually In vivo All radiation workers where 

exposure to surface or 

airborne radioactive 

contamination could give at 

least 0.1 mrem AEDE from 

occupational sources, or give 

an organ or tissue DE >5 rem 

annual.

In accordance with DOE 

Order 5480.11 (DOE 1988). 

Reporting

Workers that could receive 

0.1 rem AEDE or 5 rem ADE 

organ or tissue dose. 

Bioassay requested 

when workplace

monitoring program 

indicates >0.02 annual 

limit of intake. 

Follow-up triggered by 

positive results from 

the workplace 

monitoring program, 

positive routine 

bioassay sample, or in 

response to incidents 

involving suspected 

intakes. 

King 1990

6 months In vitro fecal Rich 1990 

Annually In vitro urine

New hire In vivo Worked at a facility where 

gamma-emitting radionuclides 

were handled.

AEDE ≥0.01 rem. 

Investigating

In vitro urine

fecal

Worked in U manufacturing or 

recovery facilities.

Worked with transuranic

materials 

Termination In vivo

In vitro 

Any employee suspected of 

having an internal exposure or 

on a scheduled monitoring 

program.
Document No. ORAUT-TKBS-0007-5 Revision No. 01 Effective Date: 01/13/2006 Page 15 of 47

Table 5-3 (Continued). Routine bioassay history summary.

Year

Typical 

frequency Type

Groups 

analyzed/sampled Investigating level Comments Reference

1995 Appropriate to 

the facility 

mission, 

potential 

uptakes. 

In vivo All radiation workers that 

enter radiological buffer 

areas or areas of higher 

radiological controls and 

are likely to receive 

intakes resulting in a 

CEDE of 0.1 rem or more. 

Type of bioassay based 

on source term. Urine 

requested when pure 

beta, uranium, or TRU 

was of interest. Feces 

requested primarily for 

uranium and TRU source 

terms. 

In accordance with DOE 

5480.11 (DOE 1988) and 

10 C. F. R. 835. 

Reporting

Workers that could 

receive 0.1 rem CEDE. 

Declared pregnant 

workers when 

embryo/fetus could 

receive 0.05 rem DE. 

Each facility has a 

specific Technical 

Basis Document for 

Internal Dosimetry. 

Andersen, 

Perry, and 

Ruhter 1995 

In vitro urine

fecal

When work 

place monitoring 

indicates 

significant 

potential for 

intakes. 

Follow-up for any 

suspected intake of 

radionuclides and to 

more accurately 

identify and 

characterize the 

amount of intake and 

excretion pattern. 

Internal doses resulting 

from all confirmed intakes 

are to be evaluated. 

Investigating

Random sampling is 

performed to demonstrate 

the adequacy of the 

radiological controls in 

limiting the internal intake 

of radionuclides. 

Employees are selected 

at random from both 

nonradiation workers and 

a radiation worker 

population.

1995-

2000 

New hire Based on screening to 

determine internal 

conditions from previous 

uptakes or to establish 

baseline for those 

continuing to work as 

radiation workers.

Termination In vivo Any employee that was 

on a scheduled 

monitoring program.

In vitro 

2001 As developed by 

individual 

facilities based 

on analysis 

tables developed 

for each 

radionuclide.

In vivo All radiation workers

In accordance with DOE 

5480.11 (DOE 1988) and 

10 CFR 835. 

Reporting

Workers that could 

receive 0.1 rem CEDE. 

Declared pregnant 

workers when 

embryo/fetus could 

receive 0.05 rem DE. 

Bioassay is 

mandatory when an 

employee or visitor is 

involved in an event 

where the internal 

uptake of 

radionuclides was 

likely to have 

occurred. 

Rielly 2001

In vitro urine 

fecal

Termination In vivo Any employee that was 

on a scheduled 

monitoring program. 

In vitro Investigating

Uranium, >1.0 μg/L 

 

In vitro activity detected 

>2σ. 

In vivo >2.33 σ. 

Default trigger levels 

exceeded. 

a. AEDE = annual effective committed dose; ERDA = Energy Research and Development Administration.

5.2.3

Formal or legal internal dose data were maintained by the DOE HSD in individual hard-copy folders 

until 1989, when all technical support service functions, including those related to internal dosimetry, 

were transferred to the INEEL prime contractor. At that time, in vitro analytical functions were 

transferred to an onsite analytical laboratory. The in vivo counting laboratory provides support directly 

through the Radiation Dosimetry and Records (RDR) organization, which administers external and 

internal dosimetry support programs. The current contractor subject matter expert reviews, validates, 

Internal Dose Records
Document No. ORAUT-TKBS-0007-5 Revision No. 01 Effective Date: 01/13/2006 Page 16 of 47

and prepares official internal dose assessments. A DOE staff member at RESL is responsible for 

oversight of INEEL internal dosimetry program functions and provides quality assurance. The RDR 

unit functions include documentation and records custodial responsibilities. In 1999, the Radiation 

Control Information Management System (RCIMS) was placed in service to support the Radiation 

Protection program, including internal dosimetry. RCIMS is a database that lists reported internal 

doses as CEDE when an individual’s dose history is prepared.

5.2.4

The following information is important to internal dose reconstruction because the worker files from 

DOE can contain a variety of internal dose information including the calculated internal doses as well 

as the in vitro and in vivo individual bioassay results. The changing regulations influenced the level of 

internal dose evaluation and documentation, but did not change the fact that all (negative as well as 

positive) bioassay data were recorded in the individual dosimetry files.

Internal Dose Regulations, Investigation Levels, and Data Codes

The information used in internal dose assessments and analytical data sheets has varied through the 

years. Table 5-4 describes coded information that could appear in records after 1989. Table 5-5 

describes internal dose information that could appear in pre-1989 records. Table 5-6 contains 

analytical nomenclature. Table 5-7 contains INEEL codes for various site areas.

Table 5-4. Internal dose assessment information after 1989. 

Coded information Description

Name & SS No. Exposed employee by name and social security number.

Asmt. Nos. This assessment number is the calendar year, i.e., (83) and a consecutive numbered 

assessment for that employee during that specific year.

Intake Date Month/Day/Year of employee intake.

Radionuclide Class & 

Amt.

Specific radionuclide followed immediately by ICRP Publication 30 solubility class symbol D, 

W, or Y (ICRP 1979). Amount in microcuries or bequerels.

CEDE rem Calculated CEDE in rem.

AEDE rem Calculated AEDE in rem.

Year Year for which the AEDE was calculated.

Organ (Max.) Organ that received the maximum dose from the specified intake.

Organ CDE rem CDE calculated for the listed organ in rem.

Employer and Exp. 

Location

Abbreviation of DOE site contractor and the plant site of exposure (can include the building 

number).

Year – Total CEDE CEDE exposures are summed for the year of intake for each employee.

Year – TL Organ CDE Organ CDE total (TL) exposures are summed for the year of intake for each employee.

Table 5-5. Internal dose assessment information before 1989.

Dose information Description

Name, Soc. Sec. No. Employee name, social security number, and (Contractor Abbreviation/Plant or Facility).

Nuclide Radionuclide symbol followed by ICRP solubility class (D, W, or Y) (ICRP 1979).

Intake Period Month and Year for single exposure or period of time by month and year in which exposure 

occurred. 

Organ (Max.) Organ that received the maximum dose from the specified intake.

Organ CDE rem CDE calculated for the listed organ in rem.

CEDE rem Calculated CEDE in rem.

AEDE rem Calculated AEDE in rem.

Year Year for which the AEDE was calculated.

Federal regulations about permissible internal dose and formal reporting requirements to the AEC, the 

Energy Research and Development Administration (ERDA), and the DOE changed periodically 

through the site’s history. While the regulations influenced the level of calculated internal dose that 

would be reported, these changes did not significantly change the analytical programs for the 

detection of internal intake. For example, the monitoring and analytical programs were designed to 

initiate, through in vitro and/or in vivo bioassay analysis, an investigation of any potential internal 
Document No. ORAUT-TKBS-0007-5 Revision No. 01 Effective Date: 01/13/2006 Page 17 of 47

intake as indicated by positive air sampling, personnel contamination, etc. Most of these recorded 

analyses did not result in detectable radionuclides. During the early years internal dose was usually 

considered separately from external dose in terms of meeting specific exposure limits, and the 

calculated dose was only reported and documented if specific dose levels were exceeded. 

Regulations required periodic urinalyses or in vivo counting or evaluation of air concentrations if the 

whole-body dose or dose commitment could exceed 300 mrem in a calendar quarter (AEC 1958, 

1963b, 1968, 1975; ERDA 1975). Changes in the reporting levels did not generally result in changes 

to the air-monitoring or bioassay programs. Each individual analytical result was documented and 

placed in individual exposure files regardless of the formal reporting requirements.

Table 5-6. Analytical information that could be in worker dose files. 

Analytical information Description

Sample No. Sample log number.

Date and Time Generally clear interpretation.

Sample Description Name of the employee, numerical sample number frequently included, additional special 

analyses performed (i.e. Sr-90, Y Separation, etc.). * Anal. For Generally gross beta and/or gross gamma. Sample aliquots evaporated for gross beta or 

counted directly in a deep well NaI scintillation counter, Specific isotopic analysis, based 

upon chemical separation or gamma spectrum also listed in this column.

Quantity Used Size of the sample aliquot – generally in ml.

U+ or K+ Trans. Note to indicate analytical correction for natural potassium and uranium.

Count Time Counting either used preset time or preset counts. Time in minutes recorded in either case.

Total Count Total number of counts recorded.

Gross Count, cpm Cpm determined by dividing total counts by time of count.

Bkgd., cpm Background cpm recorded.

Net count, cpm Gross cpm minus background cpm.

K-40 corr., cpm Additional background from K-40 identified. K-40 is not a facility occupational product; 

ignore from an internal dose reconstruction

Foreign Activity, cpm and 

dpm

Net counts corrected for K-40 and then converted to dpm based upon counter calibration. 

Uncertainty also included, which is recorded as 1σ based upon counting statistics.

dpm per a volume The activity is for the sample volume listed.

Result in μg/L These results are for uranium whether stated or not.

Table 5-7. Area codes that could be in worker dose files. 

Area code Description Area code Description

1 AEC Headquarters Bldg 20, 261, 264 TREAT

2 EBR-I 21 LX

3, 034, 035 CFA 22 GCRE

4, 042,045 MTR, TRA 23 OX

5, 053, 055 ICPP 24 ARHG

6 NRF 25 No information available

7 TAN (GE) 26, 263,265 EBR-II

8 Services 27 ML-1

9 NX (X is construction) at NRF 28 On-Site Site Survey

10 AX at TAN 29 Off-Site Site Survey

11, 113 CX at CPP 30 ANP Program at SL-1

12 EX at EBR 31 STPF

13, 133,135 SPERT, PBF 65 ECF

14 OMRE 66 Non-Security

15 SX at SPERT 67 Division of Compliance

16 SL-1 68 STEP

17, 333 MX at MTR 69 LPTF (Phillips & AEC)

18, 814,815 WP, RWMC 71 CADRE (guard force)

19, 772, 775 TAN (Phillips & AEC) 774,776 SMC

The investigation levels (the levels at which positive bioassay results triggered follow-up sampling to 

verify that detectable activity had been taken into the body) have also changed little from the early 

years to the present. Dickson (1977) established official investigation levels (Table 5-8) for acute 
Document No. ORAUT-TKBS-0007-5 Revision No. 01 Effective Date: 01/13/2006 Page 18 of 47

uptakes of radionuclides corresponding to one-tenth of the quarterly radiation standard. Later 

procedures (DOE 1988) set specific limits on those positive bioassay results that could result in

100 mrem AEDE or above as the point at which follow-up and reporting was required. With the DOE 

Radiological Control Manual (e.g., DOE 1994), this changed to 100 mrem CEDE. In addition, a 

calculated dose of 10 mrem or above would be recorded as an internal dose. These procedural limits 

did not materially affect the bioassay sampling frequency or the recording of even non-detectable 

radioactivity in bioassay samples, although the request for and number of follow-up samples and 

analyses could have been different as a function of the formal regulations in effect.

Table 5-8. Derived investigation levels (μCi) in 1977 for 

acute exposures. 

Radionuclide Inhalation (lung burden) Ingestion (total activity)

Cr-51 20 500

Mn-54 0.4 30

Co-57 2 90

Co-60 0.09 9

Zn-65 0.6 30

Zr-95 0.3 20

Ru-106 0.06 3

Sb-125 0.3 30

Cs-134 0.1 3

Cs-137 0.1 4

Ce-144 0.06 3

Pu-239 Whenever detected

Am-241 Whenever detected

Sr-90 (bone) When detected by skull counting

I-131 (thyroid) Initial content 0.27 Not provided

5.3 INTERNAL DOSE CONTROL

The radiological protection program was established to provide detection of barrier or ventilation 

failure in a timely manner. The program consisted of continuous and retrospective air and effluent 

monitoring combined with personnel and surface contamination monitoring. Detection of barrier 

failure provided the information for making decisions on evacuating personnel, increasing personnel 

protection equipment (e.g., respirators), and requesting bioassay analyses to identify internal intake. 

As a consequence of a consistent AEC/DOE policy to avoid detectable internal exposures, coupled 

with the time and technical complexity of an internal dose evaluation, the general policy at INEEL for 

internal exposures has been preventive in nature. In general, radiological materials handled at the 

site were of relatively low volume and mass and of higher activity concentration rather than metric 

tons of materials of low specific activity. The consistent NRTS/INEEL policy and practice was to 

require respiratory protection on jobs when the possibility of airborne contamination was thought to 

exist regardless of the actual measured air or surface contamination. These practices influenced the 

assumptions for dose evaluation in internal dose reconstruction.

In a related matter, the contamination control limits for the detection and control of released activity 

beyond the control boundaries were related to instrumentation capabilities and the basic philosophy of 

acceptance of detectable contamination. As a result of increased emphasis on exposures that were 

as low as reasonably achievable, some reduction in acceptable release levels was implemented. The 

contamination control limits for alpha on plant surfaces and particularly personnel were always set 

close to the MDA, such that any detectable contamination was a signal for preventative and follow-up 

evaluations and actions. Beta/gamma MDAs typically were a factor of 5 below the limits. Table 5-9 is 

a summary of control limits primarily from the CPP [Chemical Processing Plant] Health Physics 

Manual (ACC 1952) and current operating procedures.
Document No. ORAUT-TKBS-0007-5 Revision No. 01 Effective Date: 01/13/2006 Page 19 of 47

Table 5-9. Surface contamination control and MDAs. 

Period Surface location

Detection 

technique Control levels Typical MDA

1952-1960s Plant/equipment Smears 500 dpm β & 20 dpm α

per 100 cm2 

150 dpm β & 10 dpm α

per 100 cm2

Personal clothing Portable survey 

instruments

1500 dpm β & 500 dpm 

α per 100 cm2

1,000 dpm β & 500 dpm 

α per 100 cm2

Personal skin Portable survey 

instruments

Any detectable reported, 

e.g. 1,000 dpm β & 500 

dpm α per 100 cm2 

1,000 dpm β & 500 dpm 

α per 100 cm2

Shipments Smears/portable 

survey 

instrument

500 dpm β & 20 dpm α

per 100 cm2

 – smears

0.1 mrep/hr β & 500 dpm 

α per 100 cm2

150 dpm β & 10 dpm α

per 100 cm2 smears. 

0.01 mrep/hr β & 500 

dpm α per 100 cm2

1970s-present Plant/equipment 

surfaces

Smears 300 dpm β & 20 dpm α

per 100 cm2

30 dpm β & 10 dpm α

per 100 cm2

Personnel Portable survey 

instruments

Any detectable reported, 

e.g. 300 dpm β & 200 

dpm α per 80-100 cm2 

300 dpm β & 200 dpm α

per 80-100 cm2

5.3.1

The monitoring of radioactivity in the air in occupied areas was a basic element of the internal 

exposure prevention program. Beta/gamma continuous air monitors (CAMs) were used from the 

beginning of all facility and program operations in routinely occupied areas. With the exception of the 

1985 SMC Project, the primary contaminant radionuclides by activity were either MFPs or MAPs, 

which were beta/gamma emitters with maximum permissible concentrations/derived air 

concentrations (MPCs/DACs) above 1 × 10-9 μCi/cm3. TRU materials and uranium were available at 

some of the INEEL facilities, but they were nearly always well tagged with beta/gamma activity that 

allowed beta/gamma-detecting CAMs to be used to warn of possible alpha contamination or internal 

exposures.

Air Monitoring

ACC (1952) describes a CAM and three other air-sampling systems. The manual required use of a 

filter-type respirator when airborne activity exceeded 1 × 10-8 μCi/cm3 for beta/gamma activity or 

1 × 10-11 μCi/cm3 for alpha activity. An army assault-type mask was invoked when levels exceeded 

this by a factor of 10. Positive-pressure air masks were invoked if levels larger by a factor of 1,000 

occurred (ACC 1952, p IX: 4-1). 

The CAM systems provided real-time air activity evaluations (although it is not clear what the set 

points for alarms were), and fixed air samplers at several locations provided retrospective data and an 

average air concentration of beta/gamma emitters in an area or building. The fixed air filter samples 

were counted for both beta and alpha activity. Later, alpha CAMs were provided in select facilities 

where alpha contaminants could be controlling. CAMs were calibrated, and training programs for 

health physicists were established for interpreting CAM responses for such variations as situations, 

radionuclides, response times, and filter accumulations. In the event personnel were required to work 

in an area or building where known air contamination was present, respirators were worn to reduce 

internal contamination intake to levels below detectable amounts.

In general, workers were asked to submit to bioassay whenever they were in an area where a CAM 

alarm sounded. In addition, the fixed location and retrospective air-sampling system would signal the 

need for bioassay if elevated air sample results were detected. 
Document No. ORAUT-TKBS-0007-5 Revision No. 01 Effective Date: 01/13/2006 Page 20 of 47

5.3.2

DOE HSL technical reports and annual reports, coupled with facility memoranda and reports, 

documented the analytical detection capability of NRTS in the early 1950s and 1960s. Internal 

monitoring programs were in place when facility operations began in late 1951. For example, during

ANP Program-IET activity in 1956, particulate and liquid caustic filter samples of effluent were 

analyzed with gamma spectroscopy and specific chemical separations of the identified radionuclides 

(Ebersole 1956). This analytical capability to identify radionuclides by their energy spectra was 

available and used for urine and other bioassay samples. Specific separations (e.g., strontium, 

iodine) were available to quantify the radioactive components of a variety of samples of interest.

Early Technical and Analytical Capabilities at the National Reactor Testing Station

In the early days a gross beta measurement was made on an evaporated aliquot or a gamma count 

was made directly on a liquid sample, or both. Any detectable activity triggered a specific chemical 

separation analysis (generally strontium). Early analyses for plutonium generally were gross alpha 

counts on a plutonium separation; later, alpha spectroscopy was used to count and better 

characterize the results.

In 1958, the IDO HSD acquired a 256-channel gamma spectrometer with a 3- by 3-in. sodium iodide 

thallium-doped [NaI(Tl)] detector counting system for analyses of gamma-emitting radionuclides. In 

1960, the HSD obtained a 3- by 3-in. well counter for gamma analysis, which replaced gross beta 

counting as the routine analytical procedure for urine samples. AEC (1961, p. 59) states, 

“Approximately 1.5 × 10-6 μCi/ml of MFPs can be detected in 75 ml of urine in a 5-minute count which 

is about the same as was obtained with the gross beta procedure in a 20-minute count.” 

AEC (1961) outlined a basic philosophy in relation to gamma counting of bioassay samples. Gamma 

counting would be effective in all situations except for exposure to pure strontium isotopes. To guard 

against this unlikely possibility, the procedure of performing a strontium analysis for individual workers 

at risk (radiation workers) every 2 yr and at termination was established. Because of the improbability 

of finding detectable activity, all activities were to be precipitated by oxalic acid in a weak solution, 

gross beta counted, and the strontium analysis not completed unless a detectable count was obtained 

on the precipitate. A 100-ml sample of urine permitted the detection of approximately 8 × 10-8 μCi/ml

of 90Sr. 

As part of the EEOICPA coworker data program for the INEEL, images of all bioassay results prior to 

1986 were provided to the ORAU Team and the results were entered into an Access database. This 

database has over 150,000 bioassay results. The nearly 135,000 urine bioassay results fall into over 

80 categories of analytes, a few of which are the result of minor typographical differences. Figure 5-1 

shows the time course of the gross beta, gross gamma, and strontium analyses which were 

performed the most. The group beta-gamma in the early years seems to have followed the beta 

protocol and after 1964 follows the gamma protocol. Early on the gross beta assay using a 5 ml 

volume was used, but it was replaced with the gross gamma assay supplemented with strontium 

analyses typically using a 75 ml volume. In the early 1970s, the in vitro bioassay was largely replaced 

by whole body counting. In the 50s and 60s, several thousand assays were performed each year, 

well over one per monitored worker per year. The early results were reported in d/m/sample except 

for certain assays like Uranium reported in g U/l and tritium in d/m/ml. About September 1973 the 

activity units changed to μCi with a large negative exponent so that pCi would be a more natural unit. 

There is some sloppiness in the database record with the units between dpm and μCi and /sample or 

/ml so if the results or uncertainties look anomalous, that is a likely explanation. The sample size for 

alpha spectrometry went from the urine sample size (about a liter) to 1 ml in mid 1979 with no change 

in the sample uncertainty.
Document No. ORAUT-TKBS-0007-5 Revision No. 01 Effective Date: 01/13/2006 Page 21 of 47

INEEL Urine Bioassay History

10

100

1000

10000

50 55 60 65 70 75

Year

Number Samples Analyzed

beta

Gamma

beta gamma

Strontium

Figure 5-1:History of Urine Bioassay at the INEEL. 

Table 5-10 reproduces reports of the urinalysis results for 1959, 1960, and 1961 as obtained from 

Table 4 of AEC (1960), Table IV of AEC (1961), and Table XIII of AEC (1962a). These results are not 

identical but quite similar to those from the new database. The practice was to perform a gross beta 

or gamma analysis and identify specific radionuclides if the gross counts indicated activity above 

background levels. The total number of urinalyses in 1959 was 11,066; 3,524 people had radiation

badges; and 715 received external doses above 500 mrem. These numbers demonstrate that people 

provided urine samples multiple times during the year.

Table 5-11 summarizes early detection limits for various environmental water and milk samples 

between 1953 and 1965. These analytical sensitivities are applicable to analysis of body fluids or 

substances for the purpose of bioassay (AEC 1964). 

The special and routine bioassay sample analyses were performed and documented by the DOE 

analytical laboratory. Puphal (1994) reported on the procedures used for bioassay in the DOE 

Analytical Chemistry Branch beginning in 1960. These procedures were collected into a procedures 

manual in 1982 to be revised periodically (RESL 1982). Another version of the procedures after the 

analytical work was transferred to WINCO is documented (DOE 2002).

The individual analytical data were recorded, as were the specific doses for which formal evaluations 

were required. Before 1977, internal doses were assigned if the internal dose or dose commitment 

was greater than 50% of the yearly allowable body or organ doses by Manual Chapter 0524, the 

controlling regulation at the time (Aoki 1979). In 1977, the policy was changed to assign internal 
Document No. ORAUT-TKBS-0007-5 Revision No. 01 Effective Date: 

01/13/20

0

6 Page 22 of 47

Table 5-10. Urinalysis results in 1959, 1960, and 1961. 

Nuclide/ 

element of 

interest

Type 

activity

Total number performed

Statistically significant

Highest resulta Number Percent

1959 1960 1961 1959 1960 1961 1959 1960 1961 1959 1960 1961

Gross βb 8,546 8,546 30 65 15 5 0.76 0.18 17 18,820 ±632 dpm/5 ml 992 ±40 dpm/5 ml 172 ±16 dpm/ml

Gross γc 2,433 2,712 9,120 174 129 --d 7.15 4.76 4 35,972 ±310 dpm/5 ml 19,817 ±105 dpm/5 ml 1,900,235 ±876 dpm/75 ml

Co-60 β -- -- 1 -- -- 0 -- -- 0 -- -- 300 ±75 dpm/450 ml

Sr-90 β 3 105 3,248 3 0 2 100 0 2 4.12 × 10-2 dpm/ml Insignificant 183 ±8 dpm/75 ml

Sr-91 β 20 37 2 19 0 -- 95 0 0 388 ± 1.6 Insignificant 4 ±8 dpm/ml

I

e β -- 9 -- -- 2 -- 22 0 9992 ±80 dpm/ml --

Cs-137 β -- -- 40 -- -- 0 -- -- 0 -- -- 1,460 ±10% dpm/1700 ml

Ba-139 α 20 16 0 -- 80 0 -- 120 ±0.8 dpm/ml 0 --

Thf α 7 0 -- 0 0 -- 0 0 -- Insignificant 0 Insignificant

U α -- -- 4 -- -- -- -- -- 0 -- -- 10 μg/L

U-233 α 17 3 -- 1 0 -- .06 0 -- 180 ±4.0 dpm/ml Insignificant --

Pu-239 α 18 0 29 0 0 0 0 0 Insignificant Insignificant 2E-9 μCi/ml

Am-241 α 2 0 -- 0 0 -- 0 0 -- Insignificant 0 --

Totals 11,066 11,352 12,494 278 146 -- 2.51 1.29 4 

a. All except two I-131 exposures in 1961 listed under gross gamma activity are less than 10% of the permissible body burden for the radionuclide of interest. 

b. If only gross β analyses are available, the default should be Sr-90. 

c. If only gross γ analyses are available, the default should be Cs-137.

d. No data reported.

e. Iodine isotope(s) not identified in reference. Assume I-131.

f. Thorium isotope(s) not identified in reference. Assume Th-228.
Document No. ORAUT-TKBS-0007-5 Revision No. 01 Effective Date: 01/13/2006 Page 23 of 47

Table 5-11. Detection limits applicable to 

environmental sample analyses (1953 to 1965). 

Type of sample Radiation Detection limit

Water Alpha 3E-9 μCi/ml

Beta 6E-9 μCi/mLa

Tritium 4E-6 μCi/ml

Milk Iodine-131 10 pCi/Lb

Strontium-90 1.5 pCi/L

a. Reduced in October 1962 from 1.5E-7 μCi/ml. 

b. Reduced in September 1962 from 50 pCi/L.

doses when the CDE to an organ exceeded one-tenth of the quarterly radiation protection standard. 

Replies to requests for radiation exposure history before 1979 stated that there was “no positive

exposure reported” when the dose was below the reporting levels noted above. Because individuals 

remembered receiving some internal exposure, the statement, “no positive exposure reported,” was 

determined to be misleading and was changed in 1977 to “no reportable levels recorded.” In all 

cases, copies of the bioassay results were placed in the individual’s radiation exposure file and should 

be in the individual’s dose file in the NIOSH Office of Compensation and Support Claims Tracking 

System database.

5.4 MINIMUM DETECTABLE ACTIVITIES

In compliance with the November 1998 Code of Federal Regulations requirement (10 C. F. R. pt. 835)

for the DOE Laboratory Accreditation Program (DOELAP), and based on American National 

Standards Institute (ANSI) N 13.30, Performance Criteria for Radiobioassay (ANSI 1996), both the in 

vitro and in vivo radiobioassay laboratories at INEEL received DOELAP accreditation in February 

1998. In accordance with this accreditation, MDAs and decision levels at the 95% confidence level 

(2σ) are performed. Tables 5-12 and 5-13 list current MDAs for urine and fecal sample analysis, 

respectively, along with values gleaned from historical documents. A large majority of the urine 

samples taken at the site were single voidings; 24-hr samples were used for special sampling 

purposes (i.e., follow-up samples, primarily to extend the sensitivity). The MDAs listed are those for 

the primary samples; the tables include the recommended periods for the MDA values. Table 5-14 

lists the current in vivo MDAs along with values gleaned from historical documents and the 

recommended periods for use with the MDA values.

5.5 DEVELOPMENT AND IMPACT OF WHOLE-BODY COUNTING

WBC was introduced at the INEEL in 1961. As early as 1961 one of the fundamental conclusions 

from the experience at the NRTS with in vivo and in vitro internal dosimetry analytical techniques was 

the fact that a large proportion of the internal exposures to NRTS workers was to insoluble materials. 

Radioactive nuclides (e.g. 125Sb, 110mAg, 65Zn, and 95Zr/Nb) were detected by an in vivo count and not 

in the urine. Concurrent analyses of fecal and urine excreta demonstrated the main elimination route 

to be by the feces, with so little voided in the urine as to be undetectable even in a 24-hr specimen 

(AEC 1962a; Sill, Anderson, and Percival 1964). WBC was demonstrated to detect activity as low as

0.01 μCi in a 10-min count (AEC 1962a). This detection level was several orders of magnitude more 

sensitive than the maximum permissible body burdens (MPBB) for most beta/gamma fission and 

activation products.

As a consequence, the in vivo counting program was used to count 1) all terminating employees who 

required physical exams, 2) employees suspected of having a possible internal intake, and 3) 

selected groups of individuals scheduled for semi-routine analyses by health physics supervisors 

(Sommers 1961). In 1963 approximately 1,650 counts were performed; only those activities greater 
Document No. ORAUT-TKBS-0007-5 Revision No. 01 Effective Date: 01/13/2006 Page 24 of 47

than 0.1 μCi were further quantified. This level was determined to be less than one-tenth of the 

MPBB for most of the gamma-emitting isotopes.

Table 5-12. MDAs for urine samples by period. 

Radiation/

radionuclide Period 

Typical Volume 

(ml)

Typical MDA

(dpm/sample)

Typical Dailyc

MDA (dpm/d) Reference

Gross β 1951-1953 5 86 24000 Data Sheet

1954-1960 5 93 26000 Ebersole & Flygare 1957a

Gross γ 1957-1964 75 580 10800 AEC 1961a

1965-1971 75 205 3800 Data Sheet

H-3 1972-1994 3 0.5 dpm/ml 700 AEC 1972a , AEC 1974

1995-present 3 dpm/ml

4200

Andersen, Perry, and Ruhter 1995, Rielly 

2001

Co-60 1957-1958 50 51 1400 Database

Sr-90 1953-6/14/62 75 37 700 Database

6/15/62-1970 75 20 370 Database

1971-1989 75 1.7 32 AEC 1972a, AEC 1974a

1990-present 500 min 1.9 5 Andersen, Perry, and Ruhter 1995

I-131 1957-1970 75 370 6907 Database

Cs-134 1974-present 400 2 7 Database

Cs-137 1961-present 400 410 1435 Database

Th-230 1974-present 1000 0.1 0.14 AEC 1974 a

Np-237 1974-present 1000 0.1 0.14 AEC 1974 a

U (FP) 1954-1961 0.1 1E-5 g U/Lb 14 μg U b Database.

1962-1971 0.1 5E-6 g U/Lb 7 μg U b Database.

U (KPA) 1985-present 0.2 μg/L 0.28 μg U Rich 1990

U-233/234 1979-1986 1 0.52 728 Database

1995-present 500 min 0.091 0.25 Andersen, Perry, and Ruhter 1995

U-235 1970-1979 1000 0.22 0.31 Rich 1990

1980-1994 1 0.13 180 Database

1995-present 500 min 0.084 0.24 Andersen, Perry, and Ruhter 1995

U-238 1970-1979 1000 0.22 0.31 Rich 1990

1980-1994 1 0.21 290 Database

1995-present 500 min 0.067 0.19 Andersen, Perry, and Ruhter 1995

Pu-238 1981-1984 1 0.072 100 Database

1990-1994 1000 0.13 0.18 Rich 1990

1995-present 500 min 0.049 0.14 Andersen, Perry, and Ruhter 1995

Pu-239/240 1964-1970 1000 0.93 1.3 AEC 1964a

1971-1973 1000 1.03 1.4 AEC 1972a

1974-1979 1000 0.47 0.66 AEC 1974 a

1980-1989 1 0.073 100 Database

1990-1994 1000 0.060 0.084 Rich 1990

1995-present 500 min 0.060 0.17 Andersen, Perry, and Ruhter 1995

Am-241 1977-1979 1000 0.16 0.22 AEC 1974 a

1980-1989 1 0.29 400 Database

1990-1994 1000 0.2 0.28 Rich 1990

1995-present 500 min 0.051 0.14 Andersen, Perry, and Ruhter 1995

Cm-244 1974-present 1000 0.155 0.22 AEC 1974 a

a) MDA calculated from inferred 2σ uncertainty.

b) Smallest reported value. Not MDA.

c) Based on 1400 ml daily volume and typical sample size. May need adjustment for larger or smaller sample.
Document No. ORAUT-TKBS-0007-5 Revision No. 01 Effective Date: 01/13/2006 Page 25 of 47

Table 5-13. MDAs for fecal samples by period.

Radiation/

radionuclide Period

Fecala

(pCi/sample) Reference

Co-60 1963-present 10 Rich 1990

Sr-90 1963-1994 10 Rich 1990

1995-present 1.9 Andersen, Perry, and Ruhter 1995, Rielly 2001

Cs-134 1963-present 10 Rich 1990

Cs-137 1963-1999 0.01 Rich 1990

2000-present 0.3 Bechtel BWXT 2000

Th-230 1974-present 0.0003 AEC 1974

Np-237 1974-present 0.0003 AEC 1974

U-233/234 1970-2002 0.041 Andersen, Perry, and Ruhter 1995, Rielly 2001

2003-present 0.05 Bhatt 2003

U-235 1970-2003 0.038 Andersen, Perry, and Ruhter 1995, Rielly 2001

2003-present 0.09 Bhatt 2003

U-238 1970-1994 0.5 Rich 1990

1995-2002 0.03 Andersen, Perry, and Ruhter 1995, Bechtel BWXT 2000, Rielly 

2001

2003-present 0.09 Bhatt 2003

Pu-238 1974-1994 0.03 AEC 1974

1995-2002 0.022 Andersen, Perry, and Ruhter 1995, Rielly 2001

2003-present 0.02 Bhatt 2003

Pu-239/240 1964-1973 0.4 AEC 1964b

1974-1994 0.02 AEC 1974

1995-present 0.03 Andersen, Perry, and Ruhter 1995, Bechtel BWXT 2000, Rielly 

2001, Bhatt 2002

Am-241 1974-1994 0.07 AEC 1974

1995-2001 0.023 Andersen, Perry, and Ruhter 1995, Rielly 2001

2002-present 0.04 Bhatt 2002

Cm-244 1974-present 0.02 AEC 1974

Cf-252 1974-present 0.02 AEC 1974

a. When sample size is not identified in an individual’s records, assume the activity is that excreted per day.

b. MDA calculated from inferred 2σ uncertainty.

Table 5-14. In Vivo MDAs by period. 

Radiation/ 

radionuclide Period

In vivo 

MDA (nCi)

Count 

time (min) Reference

Cr-51 1962-2000 12 10 Percival and Anderson 1962a

2001-present 32 5 Rielly 2001

Mn-54 1962-2000 5 10 Martin 1989; Grothaus 1993; Andersen, Perry, Ruhter 1995

2001-present 2.6 5 Rielly 2001 

2001-present 1.3 10 Rielly 2001

Fe-59 1962-2001 4.5 5 Rielly 2001

2001-present 1.5 10 Rielly 2001 

Co-58 1962-2000 12 10 Percival and Anderson 1962a

2001-present 2.5 5 Rielly 2001

2001-present 1.1 10 Rielly 2001

Co-60 1962-1970 12 10 Percival and Anderson 1962a

1971-1988 5 10 AEC 1972a; AEC 1974

1989 7 10 Martin 1989

1990-1992 2 (lung) Rich 1990 

1993-2000 7 10 Grothaus 1993; Andersen, Perry and Ruhter 1995

2001-present 2.5 5 Rielly 2001

2001-present 1.1 10 Rielly 2001 

Zn-65 1962-1988 12 10 Percival 1962a

1989-2000 10 10 Martin 1989; Grothaus 1993; Andersen, Perry, and Ruhter 1995

2001-present 4.9 5 Rielly 2001

2001-present 2 10 Rielly 2001

Sr/Y-90 1968-1977 70 (skull) 10 AEC 1969a; AEC 1972a; AEC 1974

1978-present 34 (skull) 10 Martin 1989; Grothaus 1993

Zr/Nb-95 1962-1988 12 10 Percival and Anderson 1962a

1989-2000 5 10 Martin 1989; Grothaus 1993; Andersen, Perry, and Ruhter 1995
Document No. ORAUT-TKBS-0007-5 Revision No. 01 Effective Date: 01/13/2006 Page 26 of 47

Radiation/

radionuclide Period

In vivo

MDA (nCi)

Count 

time (min) Reference

2001-present 2.6 5 Rielly 2001

Ru-106 2001-present 27 5 Rielly 2001

2001-present 7.6 10 Rielly 2001

Ag-110m 1962-present 12 10 Percival and Anderson 1962a

Sb-125 1962-present 14 10 Martin 1989; Grothaus 1993

I-131 1962-1989 12 10 Percival and Anderson 1962 

1990-1992 2 (thyroid) 10 Rich 1990

1993-2000 0.3 (thyroid) 10 Grothaus 1993

2001-present 3.8 5 Rielly 2001

2001-present 0.13 (thyroid) Rielly 2001

Cs-134 1989-2000 5 10 Martin 1989; Grothaus 1993; Andersen, Perry, and Ruhter 1995

1990-present 2(lung) Rich 1990

2001-present 3 5 Rielly 2001

2001-present 0.96 10 Rielly 2001

Cs-137 1962-1970 12 10 Percival and Anderson 1962a

1971-1998 5 10 AEC 1972a; AEC 1974; Martin 1989; Grothaus 1993; Andersen, Perry, 

and Ruhter 1995

1999-2000 2. (lung) 10 Rich 1990

2001-present 3.1 5 Rielly 2001

2001-present 1.9 10 Rielly 2001

Ba/La-140 1962-present 12 5 Rielly 2001 

Ce-141 1962-present 9.9 5 Rielly 2001

2001-present 3.2 10 Rielly 2001

2001-present 0.11 (lung) 60 Reilly 2001 

Ce-144 1962-2000 50 10 Martin 1989; Grothaus 1993; Andersen, Perry, and Ruhter 1995

2001-present 44 5 Rielly 2001

2001-present 15 10 Rielly 2001

2001-present 0.44 (lung) 60 Rielly 2001

Eu-152 1962-present 4 10 Rielly 2001

2001-present 0.18 (lung) 60 Rielly 2001

Eu-154 1962-present 2 10 Rielly 2001 

Eu-155 1962-present 1 10 Rielly 2001

Ga-153 1962-present 6.5 10 Rielly 2001

2001-present 0.096 (lung) 60 Rielly 2001

Hf-181 1962-present 5 10 Martin 1989; Grothaus 1993; Andersen, Perry, and Ruhter 1995

Ta-182 1962-present 12 10 Percival and Anderson 1962a

Hg-203 1962-present 12 10 Percival and Anderson 1962a

Th-230 1974-present 1,000 AEC 1974

Th-234 2001-present 1.4 (lung) 60 Rielly 2001

Np-237 1974-present AEC 1974

U-235 1993 0.2 (wound) 20 Grothaus 1993 

1962-present 0.2 (lung) Rich 1990

2001 0.11 (lung) 60 Rielly 2001

U-dep/nat 1989 3 (lung) 60 Martin 1989; Grothaus 1993 

Pu-238 1989-1998 26 (lung) 60 Martin 1989; Grothaus 1993; Andersen, Perry, and Ruhter 1995

1993 1 (wound) 20 Grothaus 1993

1999-2000 30 (lung) Rich 1990

2001 54 (lung) 60 Rielly 2001

Pu-239/240 1971-1993 30 100 AEC 1972a

1974-1988 74 (lung) 100 AEC 1974

1989-1988 80 (lung) 60 Martin 1989; Grothaus 1993; Andersen, Perry, and Ruhter 1995 

1993 2 (wound) 20 Grothaus 1993

1990-2000 30 (lung) Rich 1990

2001-present 140 (lung) 60 Rielly 2001 

Am-241 1989-1999 0.6 (lung) 60 Martin 1989; Grothaus 1993; Andersen, Perry, and Ruhter 1995

1993 0.1 (wound) 20 Grothaus 1993

1990-2000 0.2 (lung) Rich 1990

2001-present 0.14 (lung) 60 Rielly 2001

The 1963 annual summary report (AEC 1964) describes the year’s follow-up for the WBC program as

presented in Table 5-15. As shown in the table, many of the individuals were counted multiple times. 

The maximum activity detected provides an upper bound on how large an activity could be found in 

someone in earlier years before the whole-body counter was operational. 
Document No. ORAUT-TKBS-0007-5 Revision No. 01 Effective Date: 01/13/2006 Page 27 of 47

Table 5-15. Summary statistics from the 1963 WBC program. 

Radionuclide Times reported Number of individuals Maximum activity (μCi)

Cr-51 15 10 1.2

Co-60 / Fe-59 848 387 1.5

Mn-54 98 51 0.16

Co-58 62 50 0.03

Zn-65 505 171 1.20

Zr-/Nb-95a 427 232 1.66

Ru-103-106b 93 75 0.22

Ag-110m 583 186 0.93

Sb-122 2 2 0.08

I-131 110 82 5.0

Cs-134 361 168 0.14

Cs-137 2332 573 1.32

Ba-/La-140 90 51 0.07

Ce-141-144 59 49 0.16

Ta-182 50 36 0.02

Hg-203 28 6 0.16

Pa-233 13 10 0.48

Np-239 1 1 1.68

Sb-125 3 3 0.1

Mo-Tc-99 8 5 0.72

I-132 8 7 <0.1

I-133 3 3 <0.1

Te-132 6 6 <0.1

Hg-197 7 3 0.7

a. Consider Zr-95.

b. Consider Ru-106.

5.6 SPECIFIC FACILITIES

Each of the facilities was responsible for its internal dosimetry monitoring program, which was

designed 1) to prevent and mitigate internal exposure and 2) to evaluate and document internal dose 

above the detectable limits. The following descriptions provide insight and default instructions for 

dose reconstruction if the specific operational facility is known.

5.6.1

General Electric built the first facilities at TAN in 1952 for the ANP Program, which was active during 

the 1950s and early 1960s before ANP was determined to be impractical. The initial mission was to 

develop reactors for aircraft propulsion. The large facilities built for this program have been used for a 

number of subsequent INEEL projects. Approximately 25 different reactor concepts and experiments 

have been conducted at this location, which features large hot cells, maintenance shops, water 

storage pools, and waste management areas (Stacy 2000).The IETs in the 1950s were conducted 

under area controls and radiological monitoring surveillance. The reactors operated in the open, and 

each test involved the release of large quantities of short-lived radioactive fission product gasses and 

volatiles. However, workers were protected by enclosures (e.g., control point buildings) and constant 

monitoring for identification of unanticipated exposures.

Test Area North

Following termination of the IET programs, the TAN facilities were used to handle, inspect, store, and 

prepare for disposal the materials from unplanned reactor excursions. Major reactor components of 

the damaged SL-1 reactor were examined in the large hot cell and prepared for ultimate disposal. 

Fuel from the damaged Three Mile Island (TMI) reactor was brought to and stored in the large water 

pool facility, where it was examined and prepared for permanent dry storage. 
Document No. ORAUT-TKBS-0007-5 Revision No. 01 Effective Date: 01/13/2006 Page 28 of 47

The operation of the hot cells, storage basins, and waste treatment facilities involved aged MFPs 

(primarily 137Cs and 90Sr/Y) with periods when 95Zr/Nb and 144Ce were present. Activation products 

were also encountered—primarily 60Co. Alpha emitters (uranium isotopes and TRU radionuclides) 

were present, but their ratios were at least 50:1 beta:alpha.

Operation of the 15 experimental reactor facilities in this area of the site resulted in short-lived fission 

products in addition to longer-lived MFPs and some MAPs. A few reactor experiments involved 

operating the reactor to destruction, with the attendant breach of containment and potential internal 

exposures.

5.6.1.1 Specific Manufacturing Capability Project

The SMC Project began late in the site history (1985) at the ANP Program site in the large hangar 

facility. The program uses DU to produce armor packages for the U.S. Army M1-A1 and M1-A2 tanks 

(Stacy 2000). During DU parts fabrication, small quantities of finely divided uranium metal and oxides 

present inhalation and ingestion potential, as indicated by routine positive personnel bioassays. 

Air monitoring is the primary method used at the SMC Project for evaluating the potential for exposure 

to airborne DU. Fixed-head air sampling throughout the plant, supplemented by CAMs, provides the 

routine information to evaluate the effectiveness of control programs and indicate potential internal 

intake. Exposures to concentrations above 0.1 DAC generally will indicate the use of respiratory 

protection and require bioassay follow-up (King 2001).

The radionuclides of concern at SMC are the isotopes of DU listed in Table 5-16. The mass 

percentages, relative activities in picocuries per microgram, and the total picocuries per microgram

are based upon Integrated Modules for Bioassay Analysis NIOSH default values.

Table 5-16. Mass and activity ratios of SMC DU isotopes.

a 

Isotope Mass % Activity %

Relative activity 

(pCi μg-1)

Total 

(pCi/μg)

U-238 99.8 83.42 0.3354 0.402

U-236 0.0031 0.05 0.0002

U-235 0.20 1.07 0.004

U-234 0.001 15.46 0.062

a. In addition to the uranium isotopes, two beta-emitting radionuclides are found in 

DU. Due to short half-lives, Th-234 and Pa-234m reach equilibrium with the 

U-238 parent within about 6 months of billet casting. Beta monitoring is an 

important part of radiological survey and measurements, even though these 

radionuclides do not contribute significantly to internal doses from intakes of DU.

Inhalation Absorption Type: Respirable particulates associated with SMC Project operations are 

probably a mixture of metal and metal oxides. The actual exposures are undoubtedly to mixtures of 

absorption types. During the 18 yr of operation, much bioassay data has been collected on a large 

number of individuals. The overall elimination patterns are consistent with type M, but probably are a 

mixture of all types. It could be too simplistic to assume a pure absorption type when the chemical 

form is not known for certain. The dose reconstructor should assume either type M or type S to 

maximize the dose to the organ of concern. Exposure to significant quantities of type F uranium at 

the SMC Project is not considered credible.

Particle Size: Detailed particle size analyses of representative samples from the various operations 

indicate that an activity median aerodynamic diameter (AMAD) of 2.4 μm is appropriate for typical 

SMC Project operations. This site-specific value of 2.4 μm AMAD is used for assessments of intakes 

at the SMC Project and is the default particle size distribution (Rielly 2001). 
Document No. ORAUT-TKBS-0007-5 Revision No. 01 Effective Date: 01/13/2006 Page 29 of 47

Chemical Toxicity: The threshold limit value (TLV) for inhalation of airborne concentrations of 

uranium and its compounds (independent of isotopic composition), as reported by the American 

Conference of Governmental Industrial Hygienists is 0.2 mg/m3

. The TLV is the time-weighted 

average concentration for a normal 8-hr workday and a 40-hr workweek, to which nearly all workers 

may be repeatedly exposed, day after day, without adverse effect. Conversion of the mass 

concentration using the specific activity of DU of 3.81 × 10-7 mCi/mg results in a radioactivity 

concentration of 7.6 × 10-11 μCi/cm3

. The DAC for Type M uranium is 3 × 10-10 μCi/cm3, which is a 

factor of 4 larger than the chemical TLV. The SMC Project staff has always been aware of the need 

to consider the chemical toxicity of SMC Project DU exposure in addition to the radiological limit.

Natural Background Uranium Excretion: Urine samples submitted by SMC Project nonradiation 

workers in 1987, 1994, and 1998 were assumed to represent nonoccupational elimination of the SMC 

Project worker population. The results ranged from 0.04 to 0.33 μg/L with wide fluctuations in 

individual measurement, some as high as 1.0 μg/L (King 2001). The average reported uranium 

concentration was 0.157 ±0.109 μg/L at 1σ uncertainty. Therefore, 0.16 μg/L is used as the 

nonoccupational component of uranium excretion for SMC Project workers, and is subtracted from 

each urine result before assessment of occupational internal dose. The bioassay results in the worker

files reflect the subtraction of 0.16 μg/L from the value determined in the laboratory bioassay result.

5.6.2

The ICPP, now known as INTEC, consists of a complex of high-enriched spent fuel storage basins, 

fuel dissolution and uranium extraction processing facilities, a high-level liquid waste storage tank 

farm, high-level waste calcining processes, and associated analytical and support capabilities. The 

ICPP was a process facility for the recovery of high-enriched uranium from spent fuels from a variety 

of national and a few foreign reactors. Because high-enriched uranium was the product, the process 

vessels had to be small for criticality control. Rather than being a plant with large canyon construction 

and complete remote control and maintenance, the ICPP processes were remotely controlled but 

contact maintenance was required; that is, maintenance personnel entered process cells and repaired 

equipment by hand. The process equipment in the cells (walls of 5-ft-thick high-density concrete) 

were decontaminated by flushing and rinsing with concentrated acids and complexing agents before

entry by health physics and maintenance personnel. These occasional operations were well planned, 

but they had high potentials for internal exposures. 

Idaho Chemical Processing Plant

Most internal doses experienced at the ICPP were from accidental releases. Table 5-17 lists unusual 

and episodic events that have occurred at the ICPP.

5.6.2.1 High-Enriched Spent Fuel Storage

The original spent fuel storage facility (CPP-603) was a 1.5-million gal, three-basin, 20-foot deep, 

unlined concrete water pit that operated from 1950 until 1984. Because the basin was unlined, the 

use of demineralized water was not feasible due to corrosion of the bare concrete. In the late 1970s, 

ion-exchange columns and a sand filter were installed for water cleanup.

The basin was used for the storage of aluminum-clad, stainless-steel-clad, zirconium-clad, and 

sodium-bonded stainless-steel fuels. During 34 yr of operation, cases of cladding leaks resulted in

MFP contamination of the basin water. Breaks in the stainless-steel cladding of sodium-bonded

elements allowed a sodium–water reaction that increased the dispersal of fission products to the pool 

water. Pool water contamination up to 0.15 μCi/ml (Rich et al. 1974) existed for periods during which 

cleanup methods were developed. As a result, during the late 1950s through the 1970s, air activity 

above the pool area was routinely measured at 10 to 25% (Rich et al. 1974) of the RCGa [1 x 10-9
Document No. ORAUT-TKBS-0007-5 Revision No. 01 Effective Date: 01/13/2006 Page 30 of 47

Table 5-17. Notable airborne incidents at ICPP. 

Date Incident

Radionuclides

released Internal dose discussion Reference

05/15/57 Iodine release to Y-Cell I-131 Y-Cell modifications resulted in 8 

personnel receiving minor thyroid doses 

in the range of 600 mrem.

Vance 1957

03/20/58 Iodine release at ICPP I-131 Radioactive iodine spread through 

makeup area to operating corridor. 

Thyroid intake to several health physics

technicians and operators in the 40-μCi 

range.

Rich 1958

Hayden 1958

10/16/59 Criticality accident - in 

shielded process system

Short-lived noble 

gases and I-131, 

-132, -133, etc.

Short-lived radioactive gases released to 

plant areas; internal doses reported as 

minimal.

Ginkel et al. 

1960

01/25/61 Criticality accident - in 

shielded process system

Short-lived noble 

gases & I-131, 

-132, -133, etc.

Short-lived radioactive gases released 

through process off-gas system to 76-m 

stack. Internal doses reported as 

minimal.

Paulus et al. 

1961

01/72 Release of ~1.0 Ci Ru-106 

from ICPP main stack

Ru-106 No internal doses detected. ERDA 1977

11/17/72 ICPP mass-spectrometry 

Pu contamination Incident

Pu-238, -239 An exposure incident involving about a 

dozen personnel resulted in 50-yr

exposure lung doses ranging up to about 

4 rem.

Wenzel 1973

Wenzel 1974

10/17/78 Criticality accident in 

shielded process system

Short-lived noble 

gases

Short-lived lived radioactive gases 

released through process off-gas system 

to 76-m stack. Internal doses reported as 

minimal.

Casto 1980

11/85 N-Cell Pu Uptake Pu-238 Internal exposures were far below DOE 

exposure limits, but showed a weakness 

in the radiological control program.

Henry and 

Slagle 1985

10/30/88 Release of ~0.2 Ci of 

Ru-106 from ICPP main 

stack

Ru-106 No internal doses detected. Hoff, Mitchell, 

and Moore 

1989

μCi/ml for soluble 90Sr (AEC 1968)] and was one of the very few operations at INEEL in which 

operators were allowed to work in fractional MPCa levels without respiratory protection for several 

hours a day. Routine bioassay sampling was increased for those personnel. The primary 

contaminants were aged MFPs, primarily 90Sr/90Y and 137Cs. For workers with established work 

histories in the Building 603 storage facility for extended periods, and if specific bioassay analyses are 

either not available or insufficient, a claimant-favorable default intake of 250 RCGa-hr/yr should be 

assumed (i.e., 1,000 hr/yr at 0.25 RCGa): 

20,000 cm3

/min × 60 min/hr × 1,000 hr/yr × 0.25 x 1×10-9 μCi/cm3 × 1×106 pCi/μCi = 3×105 pCi/yr. 

Based on the activity mix for stainless-steel fuel after 3 yr of decay provided in Table 5-18, an annual 

intake of 54,000 pCi of 147Pm, 48,000 pCi of 144Ce, 30,000 pCi of 137Cs, 26,000 pCi each of 90Sr and 

90Y, 9,300 pCi of 106Ru, 90 pCi of 239Pu, and 33 pCi of 234U can be assumed to account for over 95% 

of the dose. The remaining ~56,000 pCi is spread over many nuclides and causes less than 5% of 

the dose.

In 1984, the Fluorinel Dissolution Process and Fuel Storage Facility (FAST) was completed. It 

features a large stainless-steel-lined pool with ion-exchange cleanup systems and other features for 

improved contamination control and the reduction of chronic internal exposure potential.
Document No. ORAUT-TKBS-0007-5 Revision No. 01 Effective Date: 01/13/2006 Page 31 of 47

Table 5-18. Radiologically significant radionuclides for ICPP processed fuelsa. 

Nuclide Half-life

Absorption 

type

Aluminum-clad fuel

(decayed 1 yr)

Stainless steel-clad fuel 

(decayed 3 yr)

Zirconium-clad fuel

(decayed 5 yr)

Relative 

activity

Percent 

inhalation dose

Relative 

activity

Percent 

inhalation dose

Relative 

activity

Percent 

inhalation dose

Sr-90 28.78 yr F 2.4E-02 13.1 8.6E-2 14.1 2.0E-1 6.3

Y-91 58.51 d S 2.6E-02 1.1 9.7E-6 0.0 8.0E-10 0.0

Zr-95 63.98 d S 4.0E-02 1.2 3.3E-5 0.0 5.6E-9 0.0

Ru-106 368.2 d S 2.3E-02 5.5 3.1E-2 5.7 4.5E-3 0.2 

Cs-137 30.07 yr F 2.5E-02 1.2 9.3E-2 3.2 2.1E-1 1.4

Ce-144 284.3 d M 3.0E-01 47.4 1.6E-1 18.4 2.2E-2 0.5

Pm-147 2.623 yr M 5.6E-02 1.3 1.8E-1 3.2 8.4E-2 0.3 

U-234 245500 yr S 2.9E-09 0.0 1.1E-4 4.0 7.2E-8 0.0

Pu-238 87.71 yr M 1.3E-04 26.1 7.6E-6 1.2 3.0E-3 90.0

Pu-239 24110 yr M 4.1E-07 0.1 3.0E-4 50.0 1.6E-6 0.1 

Totalb 97.0 99.0 98.4

Mass fraction U-234 to total U 3.4E-5 5.3E-5 1.4E-4

a. Assume exposure to aluminum-clad fuel activity from the beginning through 1970, zirconium-clad fuel activity from 1971 to present, and 

stainless-steel-clad fuel activity when there is indication that the Pu-239 activity exceeds the Pu-238 activity.

b. The total percent of the inhalation dose is less than 100% because other radionuclides not included in the table contribute small 

amounts of dose. 

5.6.2.2 High-Level Wastes

The high-level waste storage tank farm consists of a series of 500,000-gal underground 

stainless-steel tanks, each of which is in a reinforced concrete bunker with alarmed sumps. Several 

major underground spills occurred in the tank farm, primarily at stainless steel to mild steel 

connections. These spills resulted in extremely high levels of contaminated soil that were removed

with remote equipment. These operations had high release potential, and were planned and 

executed with personnel protective equipment and monitoring. The contaminated soil was kept wet 

when being handled to minimize airborne contamination. Both the in vitro and in vivo data are 

documented in the personnel dosimetry files.

5.6.2.3 High-Level Waste Calcination

High-level waste calcination operations, which began in 1963, consisted of the production of a 

high-temperature calcine by spraying high-level liquid waste into a fluidized bed of calcine. The nitric 

oxides were vented up the stack following high-efficiency filtering, which left the radionuclides high-fire 

coated on calcine granules. The potential for release of high-level MFP, MAP, and TRU particulate 

activity was significant, but it was recognized and monitored with an extensive array of air and 

radiation area monitors. Facility operators were placed on routine as well as special bioassay 

schedules.

5.6.2.4 Process Analytical Facilities

The Remote Analytical Facilities and lower-level Process Sample Analytical Laboratories analyzed 

samples critical to process controls. The potential for radioactive material release and internal 

exposure was significant in these facilities. Some internal exposure incidents occurred in the 

laboratories from loss of control of process samples in hoods and other causes. These incidents 

frequently involved loss of control of TRU materials during analytical procedures in hoods or benchtop confinement-type operations.

5.6.2.5 Spent Fuel Processing and ICPP Most-Limiting Radionuclides

High-enriched fuel processing, which began in 1953, consisted of nitric acid dissolution of 

aluminum-clad elements, electrolytic nitric acid dissolution of stainless-steel-clad elements, 

hydrofluoric acid dissolution of zirconium-clad elements, and nitric acid dissolution of graphite fuel 
Document No. ORAUT-TKBS-0007-5 Revision No. 01 Effective Date: 01/13/2006 Page 32 of 47

following graphite burning. Uranium was extracted from the dissolved elements (using variations of 

the plutonium–uranium extraction process, which produced a uranyl nitrate product). The first-, 

second-, and third-cycle raffinates contained high levels of MAPs and MFPs along with TRU

radionuclides. These products, in highly corrosive matrices, were difficult to contain in the 

confinement barriers of the process piping, process off-gas, etc., and resulted in routine leaks and 

spills. Most of the leaks were confined to process cells, but they occasionally occurred in occupied 

process control and equipment areas. For this reason, the extensive network of CAMs and radiation 

area monitors (RAMs) was essential for timely detection of loss of confinement. 

Source terms for the various ICPP operations are based on the types of fuel that were reprocessed. 

All of the reprocessed fuels were highly enriched with 235U enrichments ranging from 50% to 93%. 

Most of the ICPP reprocessing involved aluminum-clad fuels from test reactors, stainless-steel clad 

fuels from the Experimental Breeder Reactor II (EBR-II), and zirconium-clad fuels from various 

reactors. There were other minor campaigns for processing minor amounts of fuel such as graphite 

fuel from the ROVER (space nuclear propulsion) program. In addition, some fuels that had no burnup 

or very little burnup were reprocessed in a hands-on operation known as Custom Processing. 

To minimize the radiological safety hazard due to the relatively volatile halogens, the fuel was 

normally decayed a minimum of 120 d before shipment to ICPP for processing. Processing of fuels 

often did not occur until years later. Because of this relatively long decay time, many of the short 

half-life radionuclides experienced considerable decay, leaving the actinides to make up a larger 

percentage of the total radionuclide inventory of the processed fuel. To give an indication of the 

radionuclide inventory in fuels that were processed at the ICPP, the ORIGEN2 code (Croff 1980) was 

used to determine the decay of previously calculated fuel inventories (Wenzel 2000) to typical decay 

times before processing. Aluminum fuel, based on fuel from the Advanced Test Reactor (ATR), was 

decayed 1 yr. Stainless-steel fuel, based on fuel from EBR-II, was decayed 3 yr. Zirconium fuel, 

based on various reactors, was decayed 5 yr. These decay times represented the minimum decay 

times that occurred before processing of the fuels. Table 5-18 lists the radiologically significant 

nuclides and the percent of the inhalation dose from inhaling activity from these three types of fuel. 

Several short-lived daughters (Y-90, Rh-106, Ba-137m, and Pr-144) have over 1/3 of the activity 

which will be followed in IMBA calculations and are not shown. The inhalation dose percentages were 

calculated using ICRP Publication 68 dose conversion factors for AMADs of 5 μm (ICRP 1995). 

Aluminum fuels were processed between 1953 and 1986, stainless-steel fuels primarily between 1977 

and 1981, and zirconium fuels between 1972 and 1988 (Staiger 2003). There are relatively long 

half-life fission products that persist for the ICPP source terms. In most cases, the source terms were 

well tagged with beta-emitting radionuclides, which allowed beta/gamma-detecting CAMs to be used 

at ICPP with the realization that they would also warn of possible alpha contamination or internal 

exposures. 

Table 5-18 contains too many radionuclides for efficient dose reconstruction. Rather than include all 

of the radionuclides in the default summary table for missed dose (Table 5-24 later in this document), 

only 90Sr, 137Cs, 144Ce, and 238Pu are included for aluminum and zirconium fuels. For stainless-steel 

fuels, the 238Pu is replaced by 239Pu. Cesium-137 was selected because it is most commonly reported 

in the in vivo results rather than for its dose contribution. For the pre-1960 period, 91Y and 95Zr are 

included, but after that the fuel being processed had decayed sufficiently so that these 58- and 64-d 

half-life radionuclides would have decayed away.
Document No. ORAUT-TKBS-0007-5 Revision No. 01 Effective Date: 01/13/2006 Page 33 of 47

One exception to this planned fuel aging was the radioactive lanthanum (RaLa) process, which 

operated in L cell of the 601/602 process building from February 1957 to 1963. This process was 

designed to extract RaLa from green fuel from the Material Test Reactor (MTR) with as little decay as 

manageable (less than 2 d). Fuel was removed from the MTR, transported about 2 miles to the ICPP 

in a heavily shielded transport container by a straddle carrier, immediately dissolved, and the barium 

element was extracted. The 140Ba/La product was shipped immediately to Los Alamos National 

Laboratory. This process released large quantities of volatile radioactive iodines, which have a higher 

potential for escaping confinement systems. Several significant internal exposure incidents occurred 

in which 131I, 132I, and 133I thyroid intakes occurred before personnel could respond to CAM alarms and 

take protective or corrective actions.

5.6.3

ANL-W, based at INEEL since 1951, continues to conduct nuclear programs. Although ANL-W 

receives internal dosimetry support from the INEEL service laboratories, its radiological safety 

programs operate under the DOE Chicago Operations Office. However, over the 52 yr of ANL-W 

experience, informal program coordination has occurred so there would be technical consistency with 

the other INEEL facility internal dosimetry programs.

Argonne National Laboratories-West

Nine experimental reactors under the technical direction of ANL-W were operated at two INEEL 

locations, one on the southwest side of the site near RWMC and the others at the current location on 

the southeast side of the site. Early reactor operations included physics critical experiments; power 

production; routine unmoderated operation, uranium-fueled, plutonium-fueled, and breeder reactor 

designs; and self-destruct experiments. As listed in Table 5-19, the radionuclides of concern ranged 

the spectrum of MFP, MAP, uranium, and TRU nuclides. 

Table 5-19. Radionuclides of concern and MDAs for ANL-W locations (Nielsen 1996). 

Radionuclide

Absorption 

types Sources/characteristics

MDAs

Urine

(pCi/ml)

In vivoa,

5-min count

(nCi)

In vivob,

10-min count

(nCi)

H-3 (HTO) F EBR-II Reactor Facility & Sodium 

Components Maintenance Shop

2.0

Mn-54 F, M EBR-II primary source. Levels low 

and decreasing

2.6 1.3

Fe-59 4.5 2.4

Co-58 2.5 1.3

Co-60 2.5 1.4

Sr/Y-90 F All facilities handling fission products 0.02c

Cs-134 F All facilities handling fission products 2.7 1.3

Cs-137 3.0 1.5

U-235 F, M, S Hot cell, hoods, glove boxes, waste, 

reactor fuel, research areas

0.02c 0.1

U-238 0.02c 1.6

Pu-238 M, S FCF, HFEF, ZPPR, Analytical 

Laboratory

0.02c 60.0

Pu-239 0.02c 161.0

Am-241 M FCF, HFEF, ZPPR, Analytical 

Laboratory

0.1

a. Using a Canberra WBC with a 5-min count. 

b. Using a large NaI WBC Detector with a 10-min count. 

c. Assumes a 100-ml sample.

5.6.4

The RWMC has supported NRTS/INEEL operations as a waste management complex since 1952 

and has received large quantities of TRU waste from Rocky Flats and other DOE facilities. Improved 

operations have resulted in a decrease in internal dose potential. The original disposal techniques 

(dumping waste in open trenches) were relatively vulnerable to airborne release in comparison to 

Radioactive Waste Management Complex
Document No. ORAUT-TKBS-0007-5 Revision No. 01 Effective Date: 01/13/2006 Page 34 of 47

current total-containment practices. The four major areas in the RWMC facility are the Subsurface 

Disposal Area (SDA) for permanent disposal of low-level waste and some early TRU waste (which will 

eventually be exhumed and repackaged); the 58-acre Transuranic Storage Area (TSA) for temporary 

storage, examination, and certification before shipment to the Waste Isolation Pilot Plant; the 

operations area; and an administrative area where no radioactive waste is permitted.

The comprehensive radiation protection program for RWMC includes extensive air monitoring, 

personnel monitoring, and surface contamination surveillance. Although infrequent, there have been 

instances of inadvertent intakes (there were two in 1992 and one in 1996). Therefore, bioassay is 

conducted randomly at the current time. 

Tables 5-20 and 5-21 summarize the major radionuclides in the RWMC waste inventory. TRU 

radionuclides are the primary contaminants in the TSA waste; all but 241Pu are alpha emitters. 

Because the materials are not homogenous, it should not be assumed that no detection of one 

radionuclide in the inventory invalidates detection of other radionuclides. 

Table 5-20. Radioactive waste inventory in the TSA. 

Waste type Volume (m3) Total Ci Radionuclide Concentration (Ci/m3) Percentage

Stored 

contact-handled 

TRU waste

65,000 4.06E+5 Pu-241 2.5E+00 44.1

Am-241 1.4E+00 24.7

Pu-238 9.7E-01 17.1

Pu-239 6.3E-01 11.1

Pu-240 1.5E-01 2.6

U-233 1.4E-02 0.2

Cm-244 0.8E-02 0.1

Table 5-21. Radioactive low-level waste inventory in the active pits in the SDA. 

Waste type Volume (m3) Total Ci Radionuclide Concentration (Ci/m3) Percentage

Low-level waste 75,600 3.35E+05 Co-60 4.1E+0 92.

Ni-63 3.3E-1 7.4

Sr-90 9.7E-3 0.22

Cs-137 9.7E-3 0.22

H-3 5.8E-3 0.13

C-14 8.9E-4 0.02

Because the waste by nature has all chemical and physical characteristics, the claimant-favorable 

absorption type should be assumed unless data indicate that another type fits the data better. 

Exposures in later years will most likely be from contaminated soil or corrosion products and a default 

5 μm-AMAD particle size distribution should be assumed as recommended by ICRP Publication 68 

(ICRP 1995).

5.6.5

The Waste Reduction Operations Complex includes several reactor facilities that operated from the 

1950s to the late 1960s and the Power Burst Facility (PBF) reactor, which operated from 1972 to 

1985. These currently inactive facilities are in a common control area. In addition, a low-level waste 

incinerator called the Waste Experimental Reduction Facility (WERF) burned waste from all INEEL 

facilities from 1982 to 2001. The WERF, which is undergoing decontamination and decommissioning, 

was a low-level waste incinerator, and its operations included some mixed waste treatment (Stacy 

2000). 

Waste Reduction Operations Complex
Document No. ORAUT-TKBS-0007-5 Revision No. 01 Effective Date: 01/13/2006 Page 35 of 47

The waste at WERF is in the form of burnable containers and the resultant high-fired and solidified 

ash. The radioactive wastes at the mixed waste storage facility and the reactors were the sources of 

the radioactivity inventory. The ashes were removed remotely to a glovebox and solidified in 55-gal 

drums. 

The radiological protection program included CAMs, fixed air-sampling systems, RAMs, surface and 

personnel contamination surveillance, and effluent monitors.

Because the types of radioactive materials processed at WERF varied dependent on the area 

shipping the waste to WERF, the claimant-favorable assumption is that radioactive materials came 

from zirconium fuel as processed at the ICPP (see Table 5-18, columns 7 and 8).

5.6.6

5.6.6.1 Test Reactor Area Reactors

Test Reactor Area

The MTR was the second operating reactor at the NRTS (from March 1953 to 1970). The TRA also 

hosted the Engineering Test Reactor (ETR; 1957 to 1981) and the ATR (1967 to present) along with 

six reactor-critical facilities that supported the test reactors. The TRA complex includes hot cells, a 

gamma irradiation pool facility, research laboratories, and analytical laboratories. The reactors at 

TRA, as well as the others at the INEEL, were used for testing materials, experiments, neutron 

irradiation facilities, and so forth. They were not involved in the production of plutonium or any other 

weapons materials, as were some DOE reactors. 

The uranium in the TRA reactors is enriched to 93% 235U, and the fuel is clad in aluminum. The 

predominant activation product in the cladding is 24Na, which is formed by activation of sodium in the 

aluminum. Sodium-24 has a half-life of 15 hr and emits a high-energy gamma ray (2.75 MeV). The 

inhalation dose to personnel from 24Na is insignificant in comparison to that from the fission products 

in the fuel. There are minor levels of activation products of stainless steel (

58Co, 60Co, 51Cr, 56Mn, etc.) 

in the primary coolant system (PCS) due to corrosion of the stainless-steel PCS components. The 

claimant-favorable recommendation is to use any in vivo counting data in the worker files directly. If 

applicable data is absent, the in vivo MDLs may be assumed consistent with information in 

Tables 5-14 and 5-24. 

Several factors contributed to unusual amounts of fission products in the PCS of the MTR and ETR 

during early operations. With cladding technology in its infancy, the quality of the cladding was not 

the best and fission products leaked through the cladding. Another factor was tramp fuel, which was 

a contaminant on the outside of the cladding. During reactor operation, fission products created in 

tramp uranium were released directly to the PCS. Reactor operators and the fuel manufacturer 

resolved these deficiencies over time; by the time the ATR became operational, the PCS of the ATR 

was considerably less contaminated than that of the MTR or the ETR during their early years of 

operation. The radiologically significant radionuclides for aluminum fuel processed at the ICPP in 

Table 5-18 are applicable to this PCS contamination.

The majority of radioactivity releases from the TRA reactors to areas potentially occupied by workers 

consisted of noble gases that promptly decayed to short-lived particulates. The principle dose to 

personnel from releases of noble gases was direct radiation rather than inhalation. The direct 

radiation caused instruments to alarm, which resulted in immediate evacuation of the affected areas. 

The claimant-favorable position is that there could have been some halogens and particulate 

radionuclides released along with the noble gases. The ORIGEN2 code (Croff 1980) was used to 

calculate a radionuclide inventory for ATR fuel at the time of reactor shutdown. The radionuclide 

inventory was then fractionated according to the release percentages often used to assess 
Document No. ORAUT-TKBS-0007-5 Revision No. 01 Effective Date: 01/13/2006 Page 36 of 47

hypothetical reactor accidents (AEC 1962b): 100% noble gases, 50% halogens, and 1% solids. A 

halogen-to-solid ratio of 50:1 was therefore used to determine the radionuclides that could 

significantly contribute to inhalation dose. Table 5-22 contains the relative amounts of halogens and 

particulate radionuclides from ATR fuel that contribute significantly to inhalation dose, but there are 

too many for efficient dose reconstruction. The iodine isotopes contribute the large majority (94.5%) 

of the dose from a test reactor gaseous release. Therefore, 131I was selected to be the representative 

radionuclide for missed dose (in Table 5-24), and the dose from 131I was weighted by a factor of 1.6 to 

account for the total iodine dose. When there is measured 131I for any facility, a dose reconstructor 

may choose to apply this factor of 1.6 to the 131I dose to account for short half-life iodines and may 

choose to ratio 144Ce, 89Sr, 91Y, and 95Zr activities to the 131I activity using the values in Table 5-22. 

Table 5-22. Radiologically significant radionuclides for ATR 

fuel gaseous releases. 

Nuclide Half-life

Absorption 

type

Fraction

activity

Ratio activity to 

I-131 activity

Sr-89 50.5 d S 0.0028 3.5E-02

Y-91 58.51 d M/S 0.0033 4.1E-02

Zr-95 63.98 d S 0.0036 4.4E-02

I-131 8.041 d F 0.0816

I-132 2.3 hr F 0.1298

I-133 20.8 hr F 0.2040

I-134 0.876 hr F 0.2225

I-135 6.61 hr F 0.1855

Ce-144 284.3 d S 0.0012 1.5E-02

Total 0.834 Not applicable

Another potential source of inhalation dose was from the pressurized-loop experiments in the TRA 

reactors. These loop experiments sometimes contained silver and tantalum, which became activated

and produced 110mAg and 182Ta. When a depressurizing incident occurred in the loop, these activation 

products were sometimes released to work areas. One of the major airborne incidents at TRA listed 

in Table 5-23 involved the release of 182Ta. The claimant-favorable position is to assign a missed 

dose based on the MDA for in vivo counting (see Table 5-14) for 110mAg and 182Ta.

Table 5-23. Major airborne incidents at TRA. 

Date Incident

Radionuclide(s) 

released Reference

03/28/1954 GE-ANP-1 depressurization Noble gas Sommers 1954

12/17/1958 GEH-4 rupture Noble gas + Iodine Sommers 1958

06/13/1967 GA-18-1 depressurization Ta-182 and Ta-183 Nertney et al. 1967

01/06/1967 Noble gas release at ATR Noble gas Sommers 1977

Reactor components, such as fuel elements, reactor loop components, and so forth that are removed 

from the reactor and placed in the canal are a source of contamination in the canal water. If these 

components are not cleaned adequately before they are removed from the canal, the activity on them 

can become airborne. The radiologically significant radionuclides listed in Table 5-18 for aluminum 

fuel processed at the ICPP are applicable to this TRA contamination.

5.6.6.2 Test Reactor Area Laboratories

The wing buildings of the MTR at TRA house chemistry and other laboratories. Over the years, 

experiments in the laboratories have resulted in contamination and airborne activity incidents. These 

experiments have involved various isotopes of plutonium and uranium along with other radionuclides. 
Document No. ORAUT-TKBS-0007-5 Revision No. 01 Effective Date: 01/13/2006 Page 37 of 47

The radiologically significant radionuclides listed in Table 5-18 for aluminum fuel processed at the 

ICPP are applicable for the experiments in the TRA laboratories.

Beginning in 1980, many of the laboratories were involved in various studies of TMI reactor fuel from 

the accident in March 1979. Because the TMI core had only the equivalent of 100 effective 

full-power d at the time of the accident, the TRU content would not be as high as that for a typical 

power reactor core, which is routinely operated for a year or so before the fuel is changed. Use of the 

radiologically significant radionuclides for zirconium fuel processed at the ICPP in Table 5-18 is 

claimant-favorable for evaluating inhalation dose from exposure to contamination from the TMI fuel.

5.7 DEFAULT FOR MISSED DOSE 

Based on the INEEL characteristics and circumstances, a number of missed dose default 

assumptions have been derived. Table 5-24 is a summary of these recommended defaults. 

For most of the history of the INEEL, personnel dosimeters were issued to all workers, regardless of 

work assignment, who entered the security access control points at each facility. For example, 

administrative and clerical personnel were required to wear these radiation-monitoring dosimeters 

even though they were not exposed to elevated backgrounds or internal dose potential. Dose 

reconstructors should determine the appropriate default categories from Table 5-24 as follows: If the 

worker’s file includes positive external dosimeter readings, those readings should be treated as 

radiation workers and the default internal missed dose should be applied as outlined in the table. If no 

detectable external or internal dose information is recorded, only the environmental dose should be 

included. For example, a worker at TRA in 1975 should be assigned the intakes for “All”, for “All but 

ANL-W”, and for “TRA”.

5.8 UNMONITORED WORKERS

As noted above, INEEL personnel dosimeters were issued to all workers at facilities handling 

radioactive material. Many of these workers, due to the nature of their work, would not have been 

exposed to internal activity and would not have been subjected to routine bioassay. 

The INEEL radiation protection program to detect the presence of and the spread of radioactive

contamination included: 

Areas with potential airborne contamination were monitored with alarming CAMs. 

The spread of contamination was monitored through the use of smears and monitoring 

instrumentation. 

It was the policy to screen radiation workers through urine samples and WBC for internal 

contamination many (often 4 or more) times a year.

Most of the activity encountered was well tagged with beta/gamma activity, which would have 

produced measurable direct radiation doses on a personnel dosimeter. Therefore, the probability that 

a worker received a significant unmonitored internal intake of radioactive material is very low. It is 

recommended that individuals who were not issued a personal dosimeter and have no record of 

internal dose monitoring be assigned only the environmental dose for the facility.

Construction workers often worked at the INEEL. Each construction job was evaluated to determine if 

radiation exposure or internal dose could be received. When construction work was done in an area 

with potential radiation exposure, including internal dose exposure, construction workers were 
Document No. ORAUT-TKBS-0007-5 Revision No. 01 Effective Date: 01/13/2006 Page 38 of 47

monitored in the same fashion as a radiation worker. Construction workers who were issued 

personnel dosimeters should be treated the same as facility employees who were issued personnel 

dosimeters. Construction workers who were not issued a personnel dosimeter should be assigned 

the environmental dose for the facility.

Table 5-24. Default table for missed dose.

Period Based on Site area Recommendation Basis

Startup–1960 Urine

gross β

All Calculate chronic Sr-90 intake that results in 

a urine activity of 0.4 × gross β

Typical β activity is 0.33 Sr-90, 0.33 Y-90 & 

0.33 Cs-137. Use of 0.4 is 

claimant-favorable 

Cs-137 intake = Sr-90 intake Half-lives and fission yields of Cs-137 & 

Sr-90 are approximately equal.

Pu-238 intake = 0.005 × Sr-90 intake Pu:Sr-90 ratio of 0.005 in Al fuel.

Ce-144 intake = 12.4 × Sr-90 intake Ce-144:Sr-90 ratio of 12.4 in Al fuel. 

Y-91 intake = 1.1 × Sr-90 intake Y-91:Sr-90 ratio of 1.1 in Al fuel 

Zr-95 intake = 1.2 × Sr-90 intake Zr-95:Sr-90 ratio of 1.2 in Al fuel

TRAa Annual I-131 intake= 25 μCi Assumes I-131 in thyroid at 3 days is 5 μCi

Vapor Type F, the maximum in vivo

measured in 1963. 

1961-1970 In vivo

Cs-137 

All Calculate chronic Cs-137 intake that results 

in the in vivo measurement

Sr-90 intake = Cs-137 intake when no in 

vitro measurement

Half-lives and fission yields are 

approximately equal.

Pu-238 intake = 0.005 × Cs-137 intake when 

no in vitro measurement

Pu:Cs-137 ratio of 0.005 in Al fuel. 

Ce-144 intake = 12.4 × Cs-137 intake when 

no measurement

Ce-144:Sr-90 ratio of 12.4 in Al fuel.

TRA Annual I-131 intake= 60 nCi when no 

measurement. 

Based on MDA for I-131 of 12 nCi at 3 days 

in thyroid Vapor Type F.

One acute intake of Ag-110m = 71 nCi when 

no measurement. 

Based on MDA for Ag-110m of 12 nCi at 3 

days type M. Known incident involving 

Ag-110m.

One acute intake of Ta-182 = 80 nCi when 

no measurement. 

Based on MDA for Ta-182 of 12 nCi at 3 

days type M. Known incident involving 

Ta-182.

1971-1980 In vivo

Cs-137

All Sr-90 intake = Cs-137 intake when no in 

vitro measurement

Half-lives and fission yields of Cs-137 & 

Sr-90 are approximately equal.

All but 

ANL-W 

Pu-238 intake = 0.02 × Cs-137 intake when 

no in vitro measurement 

Pu:Sr-90 ratio of 0.02 in Zr fuel. See Table 

5-18, column 9. 

Ce-144 intake = 1.05 × Cs-137 intake when 

no measurement 

Ce-144:Sr-90 ratio of 1.05 in Zr fuel. 

ANL-W Pu-239 intake = 0.003 × Cs-137 intake when 

no in vitro measurement

Pu:Cs-137 ratio of 0.003 in stainless steel 

fuel. 

Ce-144 intake = 2 × Cs-137 assigned intake 

when no measurement

Ce-144:Cs-137 ratio of 2 in stainless steel 

fuel.

TRA Annual I-131 intake= 94 nCi when no 

measurement. 

Based on MDA for I-131 of 12 nCi at 3 days

thyroid Vapor Type F weighted by a factor of 

1.6.

1981-present Bioassay All except 

ANL-W, 

INTEC & 

SMC

Calculate associated radionuclide intakes 

based on Al fuel distribution 

Distribution calculated for Al fuel. See Table 

5-18, column 3. 

ANL-W Calculate associated radionuclide intakes 

based on stainless-steel fuel distribution 

Distribution calculated for stainless steel 

fuel. See Table 5-18, column 4. 

INTEC & 

unknown 

Calculate associated radionuclide intakes 

based on Zr fuel distribution 

Distribution calculated for Zr fuel. See Table 

5-18, column 5. 

SMC Calculate associated intakes based on 

known isotope distribution

Isotope distribution, see Table 5-16.

a. The dose recommendations for specific areas, i.e., TRA, INTEC, ANL-W, are to be added to recommendations for “All”. 

b. Formerly known as ICPP.

c. When the worker’s area is not known, or if the employee worked in many different areas, use the dose recommendations for INTEC.

d. “Other area” or areas are known, just not one of the specific areas with special recommendations.
Document No. ORAUT-TKBS-0007-5 Revision No. 01 Effective Date: 01/13/2006 Page 39 of 47

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GLOSSARY

absorption

As used in this internal dosimetry section, absorption refers to the material being 

transported to fluids and other organs as well as radiation energy being imparted. 

activation

The process of inducing radioactivity by irradiation.

Atomic Energy Commission

An agency established by the U.S. Government for oversight of nuclear weapons and 

power production; a predecessor to the U.S. Department of Energy.

becquerel

A unit of radioactivity equal to 1 disintegration per second.

beta radiation

Radiation consisting of electrons or positrons emitted at high velocity from the nuclei of 

certain radioactive elements. Most direct fission products emit beta radiation. 

breeder reactor

A nuclear reactor in which the operation produces a net increase in fissionable material.

calcine

(1) The dry solid (grainy or granular) product of a chemical process of removing liquids 

from a solution. (2) The process for creating the chemical reaction that removes 

liquids from a solution.

cladding

The outer layer of material encasing a reactor fuel element (e.g., aluminum or 

zirconium). Cladding promotes the transfer of heat from the fuel to the coolant and 

contains fission products and activation products from the fissioning of the fuel.

contamination, radioactive

Radioactive material where it does not belong. 

core

That part of the reactor consisting of the fuel and some of the control elements for 

reactor operation.

criticality

An event that occurs as a result of uncontrolled assembly of fissile material capable of 

sustaining a nuclear chain reaction.

curie

A unit of radioactivity equal to 3.7 x 1010 disintegrations per second.

decontaminate

Removing a contaminant, such as a radioactive material, from an undesired location.

depleted uranium (DU)

Uranium that has undergone a process to remove one or more isotopes from the 

element. Depleted uranium is usually depleted in 235U and 234U, leaving primarily 238U.
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dosimeter

A device used to measure accumulated radiation exposure. 

dosimetry

The science of assessing absorbed dose, dose equivalent, effective dose equivalent, 

etc., from external or internal sources of radiation. 

enriched uranium

Uranium processed to increase the relative abundance of 235U. Most processing 

methods also increase the relative abundance of 234U.

fission

A nuclear transformation characterized by the splitting of a nucleus into at least two 

other nuclei and the release of a relatively large amount of energy.

fission product

Radionuclides that result from fission.

fuel reprocessing

A chemical process, usually involving several steps, that recovers 235U and other 

fissionable products from spent fuel.

gamma rays

Short wavelength electromagnetic radiation (photons) originating in atomic nuclei and 

accompanying many nuclear reactions (e.g., fission, radioactive decay, and neutron 

capture) in an energy range of 10,000 to 9 million electron volts. 

half-life

The time it takes for one-half of any given number of unstable atoms to decay 

(disintegrate).

hot cell

A specialized shielded laboratory in which radioactive materials can be handled with 

the aid of remotely operated manipulators. The walls and windows of the laboratory 

are made of materials designed to protect workers from radiation. 

ionizing radiation

Electromagnetic or particulate radiation capable of producing charged particles 

through interactions with matter.

in vitro

In glass. Outside the living body and in an artificial environment. Internal bioassay 

sampling, such as fecal samples or urine samples.

in vivo 

In the living. In the living body of a plant or animal. Bioassay sampling by whole-body 

counting. 

isotope

Nuclides having the same number of protons in their nuclei (same atomic number), but 

having different numbers of neutrons (different mass numbers). 
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millirem

A unit of radiation dose equivalent (or equivalent dose) equal to one-thousandth of a 

rem (see rem).

microcurie

A measure of radioactivity equal to one-millionth of a curie. 

mixed waste

Waste that contains hazardous and radioactive materials. 

natural uranium

Uranium that has not been through an enrichment process.

neutron

A basic particle in a nuclear reaction, electrically neutral, with nearly the same mass as 

a 1

H atom.

nuclear waste

A general term used for the byproduct unusable material from nuclear reactions 

including high-level, intermediate, low-level, mixed, and transuranic wastes. 

nucleus

That part of an atom consisting of the total positive electrical charge and most of the 

mass. 

photon

A quantum of electromagnetic energy often referred to as X-rays or gamma rays, but 

also including light and radiant heat. 

proton

An elementary atomic particle with a positive electrical charge equal numerically to the 

charge of the electron and a mass slightly greater than 1 mass unit.

quality factor

A modifying factor used to derive dose equivalent from absorbed dose.

radiation

Energy transferred through air or some other media in the form of particles or waves 

(see ionizing radiation).

radioactivity

The spontaneous emission of radiation, generally alpha or beta particles, gamma or 

X-rays, or neutrons from unstable atoms.

radionuclide

A radioactive species of an atom characterized by the constitution of its nucleus 

specified by atomic number (the number of protons), and the mass number (equal to 

the number of protons plus neutrons). 

radioactive lanthanum (RaLa) 

One of the fission products of a nuclear reaction; a lanthanum recovery process at the 

INTEC for development of weapons.
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rem

A unit of dose equivalent, equal to the product of the absorbed dose and the quality 

factor. The word derives from roentgen equivalent in man. 

shielding

Any material or obstruction that absorbs (or attenuates) radiation to protect personnel 

or materials.

spent nuclear fuel

Reactor fuel containing fission and activation products that can no longer economically 

sustain a chain reaction.

spent fuel storage basin

A pool or pit made of reinforced concrete containing water and used to store spent 

nuclear fuel. The water acts as shielding and as a coolant. 

type

Refers to the rate of absorption from lung to blood of inhaled radioactive materials and 

includes types F (fast), M (moderate), and S (slow).

transuranic (TRU) waste

Contaminated waste materials with nuclides having an atomic number greater than 92, 

a half-life over 20 yr, and activity concentration greater than or equal to 100 nCi/g. 

X-ray

Ionizing electromagnetic radiation of extranuclear (outside the nucleus) origin.

zirconium

A metallic element highly resistant to corrosion and often used to make cladding for 

nuclear fuel. It is sometimes alloyed in small amounts in the fuel itself.